LES AND RANS MODELING OF AN 84-PIN HEXAGONAL ROD BUNDLE WITH SPACER GRID

Adam R. Kraus, Elia Merzari

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

3 Scopus citations

Abstract

As part of an ongoing Integrated Research Project (IRP), detailed investigations of the flow and heat transfer characteristics of an 84-pin hexagonal rod bundle representing a potential modular gas-cooled fast reactor design have been performed. The rod bundle features P/D of 1.5 and a large central guide tube, along with simple spacer grids and an outer enclosure. The primary focus of the current work is modeling and simulation of this geometry using a range of computational techniques. These include high-fidelity Large Eddy Simulation (LES) and Reynolds-Averaged Navier-Stokes (RANS) approaches. The flow predictions are validated at Reynolds number of roughly 12000 using matched index of refraction Particle Image Velocimetry (PIV) experimental data that were also generated as part of the IRP. Many characteristics of the experimental flow are replicated in the LES and RANS runs, and a general agreement between simulation and experiment is shown. The LES shows notably better agreement with the turbulent kinetic energy downstream of the grid as compared to RANS. The experimental and high-fidelity data will be used to inform the closures for a novel multiscale methodology. The high-level goal of the work is to use the high-fidelity data to yield improved multiscale thermal analysis techniques for solving fuel performance problems of direct relevance to industry.

Original languageEnglish (US)
Title of host publicationProceedings of the 20th International Topical Meeting on Nuclear Reactor Thermal Hydraulics, NURETH 2023
PublisherAmerican Nuclear Society
Pages352-365
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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