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

Journal Article · Fri Dec 01 04:00:00 EST 1995 · Nuclear Technology

OSTI ID:170260

Inoue, Akira ^[1]; Futakuchi, Masanobu; Yagi, Makoto ^[2]; Mitsutake, Toru; Morooka, Shinichi ^[3]

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1. Tokyo Inst. of Tech. (Japan)
2. Nuclear Power Engineering Corp., Tokyo (Japan)
3. Toshiba Corp., Yokohama (Japan). Nuclear Engineering Lab.

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 {Delta}{alpha} = {minus}3.6%, {sigma} = 4.8% and {Delta}{alpha} = {minus}4.1%, {sigma} = 4.5%, respectively, where {Delta}{alpha} is the average of the difference measured and calculated values. The subchannel analysis codes are highly applicable for the prediction of a two-phase flow distribution within BWR fuel assemblies.

• OSTI ID: 170260
• Journal Information: Nuclear Technology, Journal Name: Nuclear Technology Journal Issue: 3 Vol. 112; ISSN 0029-5450; ISSN NUTYBB
• Country of Publication: United States
• Language: English

Related Subjects

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS
99 GENERAL AND MISCELLANEOUS
BWR TYPE REACTORS
C CODES
COMPARATIVE EVALUATIONS
COMPUTER CODES
EVALUATION
FUEL ASSEMBLIES
FUEL CHANNELS
T CODES
TWO-PHASE FLOW
VOID FRACTION