VOID FRACTION DISTRIBUTION IN A BOILING WATER-REACTOR FUEL ASSEMBLY AND THE EVALUATION OF SUBCHANNEL ANALYSIS CODES

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.