A model for predicting droplet entrainment at a quench front in thermal hydraulic analysis codes

M. J. Holowach, L. E. Hochreiter, J. H. Mahaffy, Fan-bill B. Cheung

Research output: Contribution to journalConference articlepeer-review

Abstract

The ability to accurately predict droplet entrainment at a quench front is required to effectively calculate the interfacial mass, momentum, and energy transfer, which characterizes nuclear reactor safety, system design, analysis, and performance. The present study proposes a model for droplet entrainment at a quench front, based on the best-understood physics related to the quenching phenomenon that is characteristic to Light Water Reactor (LWR) safety analysis. The model is developed based on a film boundary layer and stability analysis that attempts to match the characteristic time and length scales of the entrainment phenomenon. Data from fundamental tube reflood tests have been reduced and utilized for model development, while more-prototypical independent reflood test data have been used for model evaluation. This model has been implemented into the COBRA-TF (COolant Boiling in Rod Arrays - Two Fluid) transient three-field (continuous liquid, droplet, and vapor) two-phase heat transfer and fluid flow systems analysis computer code. Independent code calculations have been performed to confirm the viability of the model for a range of prototypical LWR reflood calculations.

Original languageEnglish (US)
Pages (from-to)3-14
Number of pages12
JournalAmerican Society of Mechanical Engineers, Heat Transfer Division, (Publication) HTD
Volume374
Issue number4
DOIs
StatePublished - Jan 1 2003
Event2003 ASME International Mechanical Engineering Congress - Washington, DC., United States
Duration: Nov 15 2003Nov 21 2003

All Science Journal Classification (ASJC) codes

  • Mechanical Engineering
  • Fluid Flow and Transfer Processes

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