Subchannel Analysis of CHF Experiments for Tight Lattice Core with COBRA-TF

Abstract

Reduced-Moderation Water reactor (RMWR) is an innovative light water reactor developed by Japan Atomic Energy Research Institute (JAERI). The RMWR comprises tight-lattice fuel assemblies with gap clearance around 1.0 mm for reduction of the water volume ratio to achieve a high conversion ratio. It is important to evaluate the thermal margin of the tight-lattice core. Subchannel analyses are expected to be useful to prediction of critical heat flux (CHF) and to provide valuable information to supplement thermal hydraulic experiments. In the present study, to assess the applicability of subchannel analysis for tight-lattice cores, series of tight-lattice CHF experiments performed in JAERI were analyzed with COBRA-TF code.For the axially uniform heated tight-lattice rod bundle, COBRA-TF gives good prediction of critical power for mass velocity of around 500 kg/(m 2 s), while it underestimates the critical power for lower mass velocity and overestimates for higher mass velocity. Predicted axial positions at BT corresponded with those of the experiments axially. However, the predicted subchannel position was outer channels and differed from the measured position. For the axially double-humped heated bundle, COBRA-TF gives good prediction of critical power for mass velocity of around 200 kg/(m 2 s), and overestimates for higher mass velocity.It turned out that the two-phase multiplier of friction loss have a large influences on the flow distribution among the subchannels. To improve the calculation accuracy, it is required to predict precisely the flow distribution including the prediction of pressure distribution in a tight-lattice bundle system. Ch2 Ch3,5 Ch4 Elevation, z [m] 1.8 Calculated dryout Pex = 7.2 [MPa] G = 500 [kg/(m 2 s)] T in = 282.5 [ o C] Ch2 Ch3,5 Ch4 Elevation, z [m] 1.8 Pex = 7.2 [MPa] G = 500 [kg/(m 2 s)] T in = 282.5 [ o C]