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A prototype liquid film sensor for high-temperature steam-water experiments has been developed. The sensor shape simulates a boiling water reactor (BWR) fuel rod. The pulse-echo method can be utilized to measure the thickness of the liquid film covering the sensor surface. A piezoelectric element is soldered onto the inside of the sensor casing which consists of two curved casing pieces. After the piezoelectric element is attached, the two casing pieces are laser welded together. It is confirmed that the temperature rise at the time of the laser welding does not influence soldering of the piezoelectric element. The pressure proof test shows that the sensor can be used at a high-pressure condition of 7 MPa. Simple air-water experiments are done at atmospheric pressure to confirm the liquid film thickness can be measured with the sensor. The fluctuation of the liquid film thickness is satisfactorily captured with the sensor. The minimum and maximum thicknesses are 0.084 and 0.180 mm, respectively. The amplitude of the waveform at 286• C is predicted by the calculation based on the acoustic impedance. It is expected that the sensor is able to measure the liquid film thickness even at BWR operating conditions.

Thermal-hydraulic design of the current boiling water reactor (BWR) is performed by correlations with empirical results of actual-size tests. However, for the Innovative Water Reactor for Flexible Fuel Cycle (FLWR) core, an actual size test of an embodiment of its design is required to confirm or modify such correlations. Development of a method that enables the thermal-hydraulic design of nuclear reactors without these actual size tests is desired, because these tests take a long time and entail great cost. For this reason we developed an advanced thermal-hydraulic design method for FLWRs using innovative two-phase flow simulation technology. In this study, detailed Two-Phase Flow simulation code using advanced Interface Tracking method: TPFIT is developed to calculate the detailed information of the two-phase flow. We tried to verify the TPFIT code by comparing it with the 2-channel air-water and steam-water mixing experimental results. The predicted result agrees well the observed results and bubble dynamics through the gap and cross flow behavior could be effectively predicted by the TPFIT code, and pressure difference between fluid channels is responsible for the fluid mixing.

A n evaluation method for droplet deposition coefficients, corresponding to spacer shapes. was developed in order to predict nuclear fuel bundle critical power. The method is based on a three-dimensional numerical analysis of dispersed flow around the spacer. The analysis showed that coolant droplet deposition characteristics onto fuel rods differed between grid type and ferrule type spacers, and also between fuel rods due to flow tabs on the upper end of the spacer band. The droplet deposition coefficient was modeled from the analysis results and adopted into a critical power prediction code based on subchannel analysis. Comparisoi.~ between critical powers calculated by the subchannel analysis code and those measured with fullscale test assemblies showed good agreement and improved prediction accuracy of the code.