Fuel-coolant interaction analysis for corium molten materials

R. H. Chen, M. L. Corradini

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

Abstract

The Fukushima accident suggested that molten fuel may have quenched in residual water in-vessel and/or likely ex-vessel from molten fuel-coolant interactions (FCI) mixing and quenching. To better understand such behavior, we have developed a molten fuel breakup model that considers solidification effects and we apply it to corium molten fuel mixing with water using the TEXAS FCI model. FCI phenomena with molten corium materials have been observed to be less energetic than predicted and we hypothesize this is due to local solidification effects. We consider the effect of a solid crust layer during fuel-coolant mixing in this model. This solidification model predicts the transient temperature profile and crust layer thickness of the fuel particle by solving the heat conduction equation for each Lagrangian fuel particle. This fuel particle breakup model and transient temperature profile model were incorporated into the TEXAS FCI model; so called TEXAS-VI model. This revised TEXAS FCI model was compared to the FARO L14 experiment (ISP-39) where fuel-coolant mixing and quench data have been published. The LI4 pressure history, liquid water pool temperature, and vapor temperature were found to be in good agreement with the revised model predictions. This mixing behavior will also have an impact on FCI explosion energetics.

Original languageEnglish
Title of host publicationInternational Meeting on Severe Accident Assessment and Management 2012: Lessons Learned from Fukushima Dai-ichi
Pages121-129
Number of pages9
Publication statusPublished - 2012
EventInternational Meeting on Severe Accident Assessment and Management 2012: Lessons Learned from Fukushima Dai-ichi - San Diego, CA
Duration: 2012 Nov 112012 Nov 15

Other

OtherInternational Meeting on Severe Accident Assessment and Management 2012: Lessons Learned from Fukushima Dai-ichi
CitySan Diego, CA
Period12/11/1112/11/15

ASJC Scopus subject areas

  • Nuclear Energy and Engineering
  • Safety, Risk, Reliability and Quality

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  • Cite this

    Chen, R. H., & Corradini, M. L. (2012). Fuel-coolant interaction analysis for corium molten materials. In International Meeting on Severe Accident Assessment and Management 2012: Lessons Learned from Fukushima Dai-ichi (pp. 121-129)