Numerical Codes Validation of Liquid Droplet Entrainment During Reflood via NRC/PSU RBHT Tests

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

Abstract

Liquid entrainment behavior in two-phase flow systems has been found to play a significant role in mass and heat transfer analysis. However, liquid droplets, generated through various mechanisms, are present in the vapor flow with sizes and velocities typically spanning different orders of magnitude. This leads to very complex two-phase flow behavior and poses challenge in accurate characterization of the mass and heat transfer processes. The objective of the current work is to provide a comprehensive evaluation of the liquid entrainment models incorporated in two different thermal hydraulic codes, TRACE and COBRA-TF, using the benchmark experimental data obtained from the NRC/PSU RBHT reflood tests. Detailed numerical simulation models have been developed in TRACE and COBRA-TF for the RBHT test facility. Based on extensive numerical calculations performed, it is found that both TRACE and COBRA-TF are able to capture the overall two-phase mass flow rate at the outlet of the test section. However, both codes tend to over-estimate the liquid droplet entrained in the post-dryout heat transfer regime while under-predicting the outlet vapor mass flow rate. Such model deficiencies may lead to bias in the peak cladding temperature predictions. Based on the current code validations, new physics-based sophisticated droplet entrainment models can be developed to realistically capture the underling physics involved in the liquid droplet entrainment process to further enhance the code prediction capability.

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
Pages740-750
Number of pages11
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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