TRACE code demonstration of thermal stratification in BWR suppression pool

Riichiro Okawa*, Masahiro Furuya

*Corresponding author for this work

Research output: Contribution to journalArticlepeer-review

5 Citations (Scopus)


An analytical model was developed to describe thermal stratification in a primary containment vessel (PCV) and transient thermal-hydraulics coupled with a reactor pressure vessel (RPV) using TRACE code version 5.0 patch level 4. Geometries of a dry well (D/W) and a suppression chamber (S/C) were represented by a nodalization of TRACE code to simulate multi-dimensional flow in the PCV. An additive loss coefficient (so called ‘K-factor’) was focused as a sensitivity parameter to limit flow rate in a pool. For the first step, a validation analysis was conducted against a steam discharge experiment of S/C. The TRACE result was in good agreement with the measurement and showed a thermally-stratified temperature distribution in the S/C pool. For the second step, an analysis to simulate the accident at Fukushima Daiichi Unit 3 power plant (1F3) was conducted. It was proved to be able to explain the pressure increase in the PCV at the beginning of accident by demonstrating thermal stratification in the S/C pool. Sensitivity study revealed an optimal K-factor value for a macroscopic viscous drag in a liquid phase fluid to demonstrate thermal stratification in a pool.

Original languageEnglish
Article number110357
JournalNuclear Engineering and Design
Publication statusPublished - 2019 Dec 15
Externally publishedYes


  • K-factor
  • Primary containment vessel
  • Thermal convection
  • Thermal stratification

ASJC Scopus subject areas

  • Nuclear and High Energy Physics
  • Nuclear Energy and Engineering
  • Materials Science(all)
  • Safety, Risk, Reliability and Quality
  • Waste Management and Disposal
  • Mechanical Engineering


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