Keiichi Hori

Real-time cross-sectional averaged void fraction measurements in vertical annulus gas-liquid two-phase flow by neutron radiography and X-ray tomography techniques

Abstract A Real-Time Neutron Radiography (RTNR) system and a high speed X-ray Computed Tomography (X-CT) system are used to determine the flow regime, the instantaneous cross-sectional averaged void fraction, and the time averaged void fraction in a vertical annulus flow channel. A standard optical video system is also used to observe the flow regime. The annulus flow channel is operated as a bubble column and measurements are obtained for gas flow rates form 0.0 to 30.0 1/min. The flow regimes observed by all three measurement systems via image analysis shows that the results agree well with each other. Both the RTNR and the X-CT systems show that the time averaged and cross-sectional averaged void fraction increases with increasing superficial gas velocity. Time and cross-sectional averaged void fractions determined by the RTNR system agree within 4% with those determined by the high speed X-ray CT technique.

Cross-sectional void fraction distribution measurements in a vertical annulus two-phase flow by high speed X-ray computed tomography and real-time neutron radiography techniques

Abstract A real-time neutron radiography (RTNR) system and a high speed X-ray computed tomography (X-CT) system are compared for measurement of two-phase flow. Each system is used to determine the flow regime, and the void fraction distribution in a vertical annulus flow channel with particular attention on the temporal resolution of the systems and the time behaviour of the two-phase flow. The annulus flow channel is operated as a bubble column and measurements obtained for gas flow rates from 0.0 to 30.0 l/min. Both the RTNR and the X-CT systems show that the two-dimensional void fraction distribution can be obtained. The X-CT system is shown to have a superior temporal resolution capable of resolving the void fraction distribution in an ( r , θ ) plane in 4.0 ms. The RTNR system is shown to obtain void fraction distribution in a ( r , z ) plane in 33.0 ms. Void fraction distribution for bubbly flow and slug flow is determined.

Development of an ultrafast X-ray computed tomography scanner system: Application for measurement of instantaneous void distribution of gas–liquid two-phase flow

An ultrafast X-ray computed tomography (CT) scanner has been developed. This scanner overcomes problems that occur in a transient or unsettled state, which make the conventional CT scanner inappropriate. To reduce the scanning time, this X-ray CT system uses electronic switching of electron beams for X-ray generation instead of the mechanical motion adopted by conventional CT scanners. The mechanical motion is a major obstacle to improving scanning speed. A prototype system with a scanning time of 3.6 ms was initially developed and confirmed to measure the dynamic events of two-phase flow. However, an increased scanning speed is generally required for practical use in the thermal hydraulics research field. Therefore, an advanced type which can operate under the scanning time of 0.5 ms and can measure two-phase flow with a velocity up to 4 to 5 m/s was developed.

APPLICATION OF CADMIUM TELLURIDE DETECTOR TO HIGH SPEED X-RAY CT SCANNER

Abstract An advanced high-speed X-ray computed tomography (CT) scanner system has been developed. This system consists of 60 X-ray tubes and 584 detectors. Cadmium telluride semiconductors are used as the detectors having suitable sensitivity, stability and time response. The concept of electrical switching of the electron beam is adopted instead of the mechanical motion of a conventional X-ray CT scanner to reduce the scanning time. The developed system can operate under a scanning time of 0.5 ms and gives 2000 slices per second. Detection and data acquisition should be performed during each X-ray shot period of about 16 μs.

Feasibility studies on high conversion pressurized water reactors with semitight core configurations

Preliminary design studies on high conversion pressurized water reactors have been completed, and their feasibility is considerably improved. Covered are nuclear design; thermohydraulic design, including departure from nucleate boiling experiments; fuel assembly; and reactor internals design performed from 1984 to 1985.

Pressurized water reactor fuel assembly subchannel void fraction measurement

The void fraction measurement experiment of pressurized water reactor (PWR) fuel assemblies has been conducted since 1987 under the sponsorship of the Ministry of International Trade and Industry as a Japanese national project. Two types of test sections are used in this experiment. One is a 5 x 5 array rod bundle geometry, and the other is a single-channel geometry simulating one of the subchannels in the rod bundle. Wide gamma-ray beam scanners and narrow gamma-ray beam computed tomography scanners are used to measure the subchannel void fractions under various steady-state and transient conditions. The experimental data are expected to be used to develop a void fraction prediction model relevant to PWR fuel assemblies and also to verify or improve the subchannel analysis method. The first series of experiments was conducted in 1992, and a preliminary evaluation of the data has been performed. The preliminary results of these experiments are described.

Post DNB Heat Transfer Experiments for PWR Fuel Assemblies

Nuclear Power Engineering Corporation (NUPEC) and Mitsubishi performed heat transfer experiments on post DNB (departure from nucleate boiling) for the pressurized water reactor (PWR) fuel assemblies under the sponsorship of the Japanese Ministry of Economy, Trade and Industry (METI) as one of a series of fuel assembly verification tests. Based on the obtained experimental data, a new evaluation model for the fuel rod heat transfer behavior after DNB was developed. A large safety margin, which had remained in the present thermal-hydraulic design that did not allow DNB, was confirmed by applying the developed model to the PWR plant safety analysis.

Experimental Study of Stable Droplet in Blowdown Phenomenon of Large Break LOCA

For blowdown phenomenon that would appear in primary stage of large break LOCA (Loss Of Coolant Accident), there exists downward misty flow over fuel rods. The behavior of droplets in such flow has effects on clad temperature because droplet removes heat from a reactor core by its direct contact on fuel rods and its evaporation at the surface and in steam flow. Droplets are generated from liquid film flowing with flashing on surface of super-heating clad or flow directory from upper plenum when coolant flows into fuel rods from upper plenum of a reactor vessel. Droplet flows in steam flow with integration and disintegration, and finally becomes steady droplet, i.e. droplet flows in the fuel rods without changing its diameter. For estimation of thermal hydraulic behavior of coolant, LOCA analysis code, MCOBRA/TRAC was developed. In order to estimate peak clad temperature (PCT) over LOCA phenomenon especially reflooding phase following blowdown phase, droplet diameter in actual plant condition must be clarified. We carried out the experiment to measure a diameter of stable droplet simulating actual condition of large break LOCA with two kinds of fluids: Sulfur Hexafluoride (SF6 ) and ethyl alcohol (C2 H5 OH) as gas and liquid phase, respectively. Those fluids can simulate a high pressure and temperature steam-water system of 6MPa, and 550K at the condition of the lower pressure and atmospheric temperature of 0.5MPa and 300K, respectively. The experiment was focused on investigating the Weaber number of steady droplet in the blowdown phenomenon of large break LOCA. It was clarified that the minimum distance was under 300mm, and the Weaber number was 3.1.Copyright