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

*この研究の対応する著者

研究成果: Article査読

17 被引用数 (Scopus)

抄録

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.

本文言語English
ページ(範囲)388-400
ページ数13
ジャーナルNuclear Technology
112
3
出版ステータスPublished - 1995 12
外部発表はい

ASJC Scopus subject areas

  • 原子力エネルギーおよび原子力工学

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