In a boiling water nuclear reactor (BWR), liquid film dryout may occur
on a fuel rod surface when the fuel assembly power exceeds the critic
al power. The spacers supporting fuel rods affect on the thermal-hydra
ulic performance of the fuel assembly. The spacer is designed to enhan
ce critical power significantly. If spacer effects for two-phase flow
could be estimated analytically, the cost and time for the development
of the advanced BWR fuel would be certainly decreased. The final goal
of this study is to be able to analytically predict the critical powe
r of a new BWR fuel assembly without any thermal-hydraulic tests. Init
ially, we developed the finite element code to estimate spacer effects
on the droplet deposition. Then, using the developed code, the spacer
effects were estimated for various spacer geometries in a plane chann
el and one subchannel of BWR fuel bundle. The estimated results of the
spacer effects showed a possibility to analytically predict the criti
cal power of a BWR fuel assembly. (C) 1997 Elsevier Science S.A.