A. Inoue et al., VOID FRACTION DISTRIBUTION IN A BOILING WATER-REACTOR FUEL ASSEMBLY AND THE EVALUATION OF SUBCHANNEL ANALYSIS CODES, Nuclear technology, 112(3), 1995, pp. 388-400
Void fraction measurement tests for boiling water reactor (BWR)simulat
ed nuclear fuel assemblies have been conducted using an X-ray computed
tomography scanner. There are two types of fuel assemblies concerning
wafer rods. One fuel assembly has two water rods; the other has one l
arge water rod. The effects of the water rods on radial void fraction
distributions are measured within the fuel assemblies. The results sho
w that the water rod effect does not make a large difference in void f
raction distribution. The subchannel analysis codes COBRA/BWR and THER
MIT-2 were compared with subchannel-averaged void fractions. The predi
ction accuracy of COBRA/BWR and THERMIT-2 for the subchannel-averaged
void fraction was Delta alpha = -3.6%, sigma = 4.8% and Delta alpha =
-4.1%, sigma = 4.5%, respectively, where act is the average of the dif
ference between measured and calculated values. The subchannel analysi
s codes are highly applicable for the prediction of a two-phase flow d
istribution within BWR fuel assemblies.