Evaluation of critical heat flux of tight lattice core with subchannel analysis code NASCA

Reduced-Moderation Water reactor (RMWR) is a light water breeder reactor developed by Japan Atomic Energy Research Institute (JAERI). The RMWR comprises tight lattice fuel assemblies with gap clearance of around 1.0 mm to reduce water volume ratio to achieve a high conversion ratio. It is important to estimate the thermal hydraulic safety margin of the tight lattice core of the RMWR. In the present study, the boiling transition (BT) prediction performance of the subchannel analysis code NASCA developed for the current BWR cores was assessed for series of tight lattice critical heat flux (CHF) experiments performed in JAERI. 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 1.8m. Axial power distribution was flat. With a simple subchannel model, the code overestimates the critical power in the high mass velocity region, although the predicted critical powers in the low mass velocity region were in good agreement with the experimental results. The predicted BT subchannel position differed from the measured one, since the amounts of liquid film in subchannels were not adequately predicted. To improve the prediction accuracy, a fine subchannel model was applied. The result implies that the use of the fine subchannel model can improve the critical power prediction in the high mass velocity region. It can also improve BT position prediction. (author)