抄録
It is important to evaluate the thermal margin of the Reduced-Moderation Water Reactor which consists of tight lattice fuel assemblies with gap clearance around 1.0mm. Subchannel analyses may provide valuable information to supplement thermal hydraulic experiments. To assess the applicability of subchannel analysis for tight lattice cores, critical heat flux experiments for tight lattice cores were analyzed with COBRA-TF code. The test section was a 7-rod bundle with rod diameter of 12.3 mm, rod gap of 1.0 mm and heated length of 1.8 m. It was found that COBRA-TF gives good prediction of critical power for mass velocity of 400-500kg/(m^2s), while it underestimates the critical power for lower mass velocity and overestimates for higher mass velocity. With modification of interfacial heat transfer model, the difference between measured and predicted powers in the high mass velocity region was reduced.
