Critical power prediction method for core thermal-hydraulic evaluation based on liquid film flow model

In the core thermal-hydraulic design for nuclear power plants, it is important to estimate the thermal margin of fuel bundles. The subchannel analysis codes based on the liquid film flow model can predict the critical power of a fuel rod bundle with high accuracy. However, it is difficult to set a subchannel analysis code in a core thermal-hydraulic code, because much time is required to calculate the critical power of an entire core. Therefore, the new critical power prediction method based on the liquid film flow model has been developed for a core thermal-hydraulic code. The method was applied to a 36-rod fuel bundle of an advanced thermal reactor, and the results showed good agreement with the experimental data under the rated mass flux condition. Bundle critical power can be calculated by taking the power ratio of a dryout subchannel to a fuel bundle. The method of selecting the dryout subchannel and a way to modify the flow rate are examined. The subchannel analysis code based on the liquid film flow model, SILFEED-ATR, was used. The calculation procedure and the verification of this method are reported. (K.I.)