TY - GEN
T1 - VALIDATION OF SYSTEM ANALYSIS MODULE AXIAL MIXING MODEL FOR THERMAL STRATIFICATION AND MIXING APPLICATION IN THE UPPER PLENUM OF A LIQUID METAL REACTOR
AU - Ross, Molly
AU - Bindra, Hitesh
AU - Zou, Ling
N1 - Publisher Copyright:
© 2023 Proceedings of the 20th International Topical Meeting on Nuclear Reactor Thermal Hydraulics, NURETH 2023. All rights reserved.
PY - 2023
Y1 - 2023
N2 - Liquid metals are being investigated as reactor coolants due to their high thermal conductivity and effectiveness at high temperatures. System analysis codes are being developed to model advanced, non-light water reactors, such as liquid metal-cooled reactors. One such code being developed is System Analysis Module (SAM) at Argonne National Laboratory, which contains capabilities for core neutronics, heat transfer, and fluid flow. System fluid dynamics are solved using a hybrid finite element/volume method with the geometry modelled as 0D or 1D components. Low-Prandtl number coolants, such as liquid metals, can result in thermal stratification, which can affect heat removal in the upper plenum of the reactor. Stratification and thermal mixing are modelled within SAM using a modified energy conservation equation which includes terms to account for mixing from local flow velocity, geometry, and buoyancy effects. While this mixing model has been validated against high-fidelity CFD models, more rigorous validation is needed. An established, scaled liquid metal test facility that uses gallium as a surrogate fluid has been constructed to investigate the stratification and thermal mixing of low-Prandtl number fluids in the upper plenum of a liquid metal-cooled reactor. This facility is used to validate the 1D axial mixing model and 0D mixing models in SAM using multiple geometries and mixing parameters. The resultant temperatures are then compared with the experimental temperatures to validate the mixing model and to better correspond mixing parameters with various flow scenarios.
AB - Liquid metals are being investigated as reactor coolants due to their high thermal conductivity and effectiveness at high temperatures. System analysis codes are being developed to model advanced, non-light water reactors, such as liquid metal-cooled reactors. One such code being developed is System Analysis Module (SAM) at Argonne National Laboratory, which contains capabilities for core neutronics, heat transfer, and fluid flow. System fluid dynamics are solved using a hybrid finite element/volume method with the geometry modelled as 0D or 1D components. Low-Prandtl number coolants, such as liquid metals, can result in thermal stratification, which can affect heat removal in the upper plenum of the reactor. Stratification and thermal mixing are modelled within SAM using a modified energy conservation equation which includes terms to account for mixing from local flow velocity, geometry, and buoyancy effects. While this mixing model has been validated against high-fidelity CFD models, more rigorous validation is needed. An established, scaled liquid metal test facility that uses gallium as a surrogate fluid has been constructed to investigate the stratification and thermal mixing of low-Prandtl number fluids in the upper plenum of a liquid metal-cooled reactor. This facility is used to validate the 1D axial mixing model and 0D mixing models in SAM using multiple geometries and mixing parameters. The resultant temperatures are then compared with the experimental temperatures to validate the mixing model and to better correspond mixing parameters with various flow scenarios.
KW - SAM
KW - Stratification
KW - Thermal mixing
UR - http://www.scopus.com/inward/record.url?scp=85202963615&partnerID=8YFLogxK
U2 - 10.13182/NURETH20-40276
DO - 10.13182/NURETH20-40276
M3 - Conference contribution
AN - SCOPUS:85202963615
T3 - Proceedings of the 20th International Topical Meeting on Nuclear Reactor Thermal Hydraulics, NURETH 2023
SP - 5487
EP - 5498
BT - Proceedings of the 20th International Topical Meeting on Nuclear Reactor Thermal Hydraulics, NURETH 2023
PB - American Nuclear Society
T2 - 20th International Topical Meeting on Nuclear Reactor Thermal Hydraulics, NURETH 2023
Y2 - 20 August 2023 through 25 August 2023
ER -