ON THE IMPACT OF PRANDTL NUMBER ON TEMPERATURE IN PARALLEL JET MIXING

John Acierno, Elia Merzari

Research output: Chapter in Book/Report/Conference proceedingConference contribution

1 Scopus citations

Abstract

Sodium-cooled Fast Reactors (SFRs) have the potential to revolutionize the future of nuclear energy due to their improved efficiency and safety measures. However, thermal striping, a phenomenon caused by fluctuations in temper-ature from mixing non-isothermal coolant streams, is a potential material safety concern that stems from the use of sodium. This phenomenon can potentially cause thermal fatigue damage in reactor upper internal structures. Although common in sodium, it can be a concern in other coolants as well. To ensure safe reactor design and operation, accurate simulations of thermal striping are crucial. To develop a thermal striping model, characterizing coolant stream mixing is the first step. In this study, we model a four-parallel plane jet geometry at various Prandtl numbers to characterize non-isothermal coolant stream mixing in a simplified upper plenum environment. To achieve accurate results, we performed Large Eddy Simulations (LES) using NekRS, an open-source spectral element CFD solver. The model was previously validated against experimental data from The University of Michigan. In this study, we will compare Power Spectrum Densities (PSD) of temperature to characterize frequency-based phenomena at different Prandtl numbers. Additionally, we will present Proper Orthogonal Decomposition (POD) of each Prandtl number to discern spatial tem-perature oscillation effects. This study aims to further our understanding of thermal striping and provide a benchmark for more cost-effective models, by studying the impact of Prandtl on jet mixing.

Original languageEnglish (US)
Title of host publicationProceedings of the 20th International Topical Meeting on Nuclear Reactor Thermal Hydraulics, NURETH 2023
PublisherAmerican Nuclear Society
Pages1332-1345
Number of pages14
ISBN (Electronic)9780894487934
DOIs
StatePublished - 2023
Event20th International Topical Meeting on Nuclear Reactor Thermal Hydraulics, NURETH 2023 - Washington, United States
Duration: Aug 20 2023Aug 25 2023

Publication series

NameProceedings of the 20th International Topical Meeting on Nuclear Reactor Thermal Hydraulics, NURETH 2023

Conference

Conference20th International Topical Meeting on Nuclear Reactor Thermal Hydraulics, NURETH 2023
Country/TerritoryUnited States
CityWashington
Period8/20/238/25/23

All Science Journal Classification (ASJC) codes

  • Nuclear Energy and Engineering
  • Instrumentation

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