To improve uranium utilization, a design study of the Reduced-Moderation Water Reactor (RMWR) has been carried out at the JAERI. In this reactor, the nuclear fission reaction is designed to be realized mainly by high energy neutrons. To achieve this, the volume of water used to cool the fuel rods is decreased by reducing the gap width between the fuel rods. For the feasibility evaluation of this reactor concept, determining the heat removal capability is very important. However, the available database was inadequate for evaluation of the heat removal capability of this type reactor core.
Using a test section simulating the proposed core design, a tight pitch arrangement of fuel rods, experiments were performed to measure the heat removal capability, i.e., the critical heat flux, under PWR thermal-hydraulic conditions. Cross-sectional views of the test section and flow channel are shown in Figs. 2-1 and 2-2. The experiments were performed with gap widths of 1.5, 1.0, and 0.6 mm between the seven heater rods in the flow channel. As experimental parameters, the flow rate and coolant temperature were changed under the same pressure conditions as in an actual reactor. Figure 2-3 shows a computer evaluation used for the design study of the actual heat removal capability. This evaluation is a comparison of the critical heat flux values calculated by a computer code and the experimental data. This figure shows that a conservative design is obtained from the computer code used in this study, since the computer code gives much smaller heat removal rates than the actual rates based on experimental data. Experiments for BWR conditions will be performed in the future. |
