Void fraction distribution in a boiling water reactor fuel assembly and the evaluation of subchannel analysis codes

Akira Inoue, Masanobu Futakuchi, Makoto Yagi, Toru Mitsutake, Shinichi Morooka

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Abstract

Void fraction measurement tests for boiling water reactor (BWR) simulated nuclear fuel assemblies have been conducted using an X-ray computed tomography scanner. There are two types of fuel assemblies concerning water rods. One fuel assembly has two water rods; the other has one large water rod. The effects of the water rods on radial void fraction distributions are measured within the fuel assemblies. The results show that the water rod effect does not make a large difference in void fraction distribution. The subchannel analysis codes COBRA/BWR and THERMIT-2 were compared with subchannel-averaged void fractions. The prediction accuracy of COBRA/BWR and THERMIT-2 for the subchannel-averaged void fraction was Δα = -3.6%, σ = 4.8% and Δα = -4.1%, σ = 4.5%, respectively, where Δα is the average of the difference between measured and calculated values. The subchannel analysis codes are highly applicable for the prediction of a two-phase flow distribution within BWR fuel assemblies.

Original languageEnglish
Pages (from-to)388-400
Number of pages13
JournalNuclear Technology
Volume112
Issue number3
Publication statusPublished - 1995 Dec
Externally publishedYes

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ASJC Scopus subject areas

  • Nuclear Energy and Engineering

Cite this

Inoue, A., Futakuchi, M., Yagi, M., Mitsutake, T., & Morooka, S. (1995). Void fraction distribution in a boiling water reactor fuel assembly and the evaluation of subchannel analysis codes. Nuclear Technology, 112(3), 388-400.