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Study of Turbulence and Pressure Recovery in the Heat Pipe Vapor Flow Using the Spectral-Element Method

  • Carolina Bourdot Dutra
  • , Tri Nguyen
  • , Elia Merzari
  • , Joshua E. Hansel

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

Abstract

Heat pipes can efficiently and passively remove heat in nuclear microreactors. Nevertheless, the flow dynamics within heat pipes present a significant challenge in designing and optimizing them for nuclear energy applications. This work aims to explore the vapor core of heat pipes through comprehensive two- and three-dimensional simulations, with a primary focus on modeling the pressure recovery observed in the condenser section. The goal is to establish improved correlations for one-dimensional heat pipe codes. The simulations are validated against experimental data from a vapor pipe documented in the literature. The turbulence model is employed in the two-dimensional simulations through the open-source spectral-element code Nek5000. This model provides insights into pressure recovery within heat pipes with low computational cost. In addition, Large Eddy Simulations (LES) are used to capture turbulent flow features in a three-dimensional vapor pipe model, utilizing the code NekRS. Using LES is crucial for comprehending the impact of laminar-to-turbulent transition on pressure recovery. A simulation framework is created to model the heat pipe's vapor core, laying the foundation for an enhanced understanding of heat pipe behavior. The ultimate goal is to improve and optimize heat pipe designs, provide data to validate lower-fidelity approaches and enhance their performance in nuclear reactors.

Original languageEnglish (US)
Title of host publicationProceedings of the 14th International Topical Meeting on Nuclear Reactor Thermal-Hydraulics, Operation, and Safety, NUTHOS 2024
PublisherAmerican Nuclear Society
Pages412-423
Number of pages12
ISBN (Electronic)9780894482212
DOIs
StatePublished - 2024
Event14th International Topical Meeting on Nuclear Reactor Thermal-Hydraulics, Operation, and Safety, NUTHOS 2024 - Vancouver, Canada
Duration: Aug 25 2024Aug 28 2024

Publication series

NameProceedings of the 14th International Topical Meeting on Nuclear Reactor Thermal-Hydraulics, Operation, and Safety, NUTHOS 2024

Conference

Conference14th International Topical Meeting on Nuclear Reactor Thermal-Hydraulics, Operation, and Safety, NUTHOS 2024
Country/TerritoryCanada
CityVancouver
Period8/25/248/28/24

All Science Journal Classification (ASJC) codes

  • Computational Mechanics
  • Safety, Risk, Reliability and Quality
  • Nuclear and High Energy Physics
  • Fluid Flow and Transfer Processes

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