VALIDATION OF THE THERMAL-HYDRAULIC MODELS IN COBRA-TF BASED ON THE NRC/PSU RBHT BENCHMARK

Yue Jin, Fan Bill Cheung, Stephen M. Bajorek, Kirk Tien, Chris L. Hoxie

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

1 Scopus citations

Abstract

In order to increase the safety margin and economic efficiency of nuclear power plants, accurate prediction of the reactor core thermal-hydraulic behavior especially during accidents is crucial. However, current thermal-hydraulic codes are still largely limited by their modeling capabilities for the post-dryout flow and heat transfer regimes. Fortunately, with the NRC/PSU RBHT experimental data recently obtained under the ARTHUR program, we are able to answer many challenging questions in post-dryout regimes and to further improve the code performance. In the current study, detailed code validation work is performed utilizing the COBRA-TF sub-channel analysis code to evaluate its liquid entrainment prediction capabilities for the two-phase flow reflood transient simulations based on the NRC/PSU RBHT benchmark. Key figures of merit (FoMs) for reflood thermal-hydraulics are evaluated including: the quench front propagation, peak cladding temperature (PCT) variations, two-phase flow mixture mass flow as well as separate phase mass flow rate. The predicted results are investigated for each of these quantities in detail to gain a better understanding of the code capabilities. A comprehensive evaluation of the models used for liquid entrainment in COBRA-TF is presented, which will be instructive and useful for future development of improved models and code upgrade.

Original languageEnglish (US)
Title of host publicationProceedings of the 20th International Topical Meeting on Nuclear Reactor Thermal Hydraulics, NURETH 2023
PublisherAmerican Nuclear Society
Pages5512-5525
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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