A wireless electrical resistance detector has been developed to measure the real-time particle volume fraction in a centrifuge for controlling a rotational velocity and particle supply rate. The purpose of this study is development of real-time particle concentration measurement technique in centrifuges because a technique for measuring the particle concentration in the centrifuges has not existed yet. The wireless electrical resistance detector was attached around a centrifuge for real centrifuge measurement experiment. This experimental centrifuge had electrodes which were covered with dielectric poly methyl methacrylate to avoid conductivity around the electrodes. The wireless electrical resistance detector was lightweight because the centrifuge condition is very severe because the rotational velocity was 3790rpm and gravity was 2000G. At first, sodium chloride solution was put in the centrifuge for the reaction stabilization of the wireless electrical resistance detector. Next, slurry of sodium chloride solution and acrylic particles with 50μm was continuously supplied in centrifuge for 10 minutes. As a result, the measured voltage was increased because the slurry was piled up in the vicinity of centrifuge inner wall. The measured resistance was increased with the lapse of time. The measured resistance coincides with an anticipated tendency of resistance. A change of particle volume fractions was measured in the centrifuge by the wireless electrical resistance detector.

Since boiling heat transfer has a high heat transfer coefficient, it has been used as a cooling technique for high-temperature bodies and it has been investigated for more than 70 years [1]. However, it has not yet been fully clarified: because the boiling phenomena are affected by many factors, such as the coalescence of bubbles, the fluid convection, the heat conduction and the physics on the contact of the gas, the liquid and the solid phase, boiling phenomena are considerably complicated. The present paper investigated an effect of pressure on boiling heat transfer mechanism by using the MEMS technology. And, boiling heat transfer enhancement in water under low pressure and low boiling temperature was examined experimentally. Steady state pool boiling experiments were conducted by using a copper thin-film and a silicon wafer for the test heater and pure water at atmospheric condition for the test liquid. The system pressure was 0.010, 0.10 and 0.15 MPa, respectively. The heaters were made of a printed circuit board and a commercial silicon wafer. The width was 7.5 mm, the length was 10 mm. The test sections were arranged for horizontal position facing upward. The test heaters had an artificial cylindrical-cavity of 0.010 or 0.040 mm in diameter; the cavities were fabricated by using the MEMS technology, i.e., wet etching technology and deep RIE. The test heaters were heated by Joule heating of d.c. current from a low-voltage high-current stabilizer. The heating rate of the heater was determined from supplied current and voltage. The temperature of the heater was obtained by referring to the measured electric resistance. The present experimental results showed the boiling bubble grew up to about 20 mm in diameter then the bubble released without coalesce of bubble under low pressure condition. Thus, the bubble coalesce was slight. From the experimental results, the gradient of boiling curve by using the copper thin-film was about 3: the heat transfer characteristic was dominant to nucleate boiling. On the other hand, the gradient of boiling curve by using the Silicon wafer (Non-cavity) was about unity: the heat transfer was dominant to heat convection of single-phase flow. According to the present observation of the boiling bubbles, the boiling heat transfer was dominant to latent heat: the ratio of the phase change and the convection was about 90 % and 10 %, respectively. The heat transfer ratio of the convection increased as the system pressure increased.

One of the important severe accident management measures in a Light Water Reactor (LWR) is water injection to the reactor core. Reflooding of the uncovered reactor core is essential to prevent total core degradation. The series of QUENCH tests have been conducted to acquire knowledge of the reflooding. A number of analyses on QUENCH tests have also been done by different computer codes for code validation and improvements. In this study, the modeling of the QUENCH-06 experiment was performed with RELAP/SCDAPSIM computer code. The input deck was modified and the SCDAP model was improved in order to represent the experimental facility more precisely. The uncertainties regarding the electrical resistance distribution in the heater rod system and the thermal properties of the shroud insulator were assessed, respectively. The main purpose of this study is to identify the models to be improved. The rather good agreement between the calculation results and the measurement was acquired than the past studies [1]. In addition, more accurate modeling of the electrical resistance and the thermal properties of shroud insulator was done and their importance was indicated. The temperature profile and oxide thickness still showed similar tendencies with the original case. Further improvements are required mainly in the heat transfer model and the oxidation model in the SCDAP code.

Heating elements are one of the important components of High Temperature Helium Loop. By means of heating elements the heating of test section of the loop to required temperature is provided even in high gas flow rates in the loop. The insulation of heating elements was manufactured of cordierite ceramics. With these elements the maximum projected temperature 900 °C could not be reached because of decreasing of electric resistance among the spirals. Therefore several tests were done in order to discover the reason of the problem. Consequently properties of new ceramic material based on corundum for manufacturing of the new heating elements were verified.

A fixed bed reactor that operates in gas-liquid co-current down flow is called Trickle Bed Reactor (TBR). It is widely used in chemical engineering. And, recently used in purification of radioactive contamination from contaminated water generated in the Fukushima Daiichi nuclear power plant. There are several flow conditions that occur in the TBR due to gas and liquid flow rate. Since mass and heat transfer rate and particles wetting depend on flow condition, it is necessary to establish the visualization techniques to understand flow condition, transition boundary and properties of gas liquid flow in TBR. In this study, authors employed the lab-scale TBR, made of 100mm inner diameter acrylic column, packed with particles of two sizes (3, 5 mm) that are used in the actual reactor. Water and air were injected from the top of the column and cross-sectional liquid distribution was captured at the bottom of the column by electrical resistance tomography (ERT). ERT is a tomographic technique that provides the cross-sectional conductivity distribution at the rate of about 50 frames per second by injecting current and measuring voltages between the 16 electrodes that are attached around the column. By analyzing the spatial and temporal characteristics of the liquid distribution obtained by ERT, it was found that particle size has only little impact on induction of pulsating flow and larger particle causes distinct pulses. Smaller particle causes blurred tiny pulses due to higher flow resistance. Larger particle (5 mm) is advantageous for pulsating flow.

In this paper, investigations of the safety of electrical cables in nuclear plant fire were described by CFD analysis and by flammability tests of power and control cables. For CFD analysis, thermal configurations of a nuclear power plant room in a fire accident was calculated by the Fire Dynamics Simulator (FDS). For flammability tests of cables, the cone-calorimeter test machine was modified to use the cable tray type specimen. The modified cone-calorimeter test machine is able to test the flammability of cables and insulation resistance from room temperature to 800°C. In the flammability test, electric power and control cables of the polyethylene sheathed cables which had been applied to the nuclear plants were adopt. The cable specimens of which length were 600mm, were fixed on a cable tray specimen. The temperature of the cable specimens was raised to from 200 to 570°C using the cone-type electrical heater. During the tests, electrical resistance and temperature was measured to evaluate safety of insulation resistance of cables. As results, the destruction of the insulation resistance of a polyethylene sheathed control cable was observed in the examinations, temperatures of which were 400°C and 500°C. In the numerical simulation using Fire Dynamics Simulator (FDS), an electrical switchgear room in nuclear plants was supposed as an evaluation model. As the fire source, fire in motor control cabinet in the switchgear room on nearby cable tray was applied to CFD analysis. From the results of numerical simulations, the heat configuration and temperature in the switchgear room was evaluated and radiation heat from fire to the cable trays was calculated. Considering these results given by flammability tests of electrical cables and CFD analysis, the safety of a switchgear room in nuclear plants was shown.

Revealing gas-liquid permeable flow phenomena in the porous media is helpful for the environmental impact assessment of geological disposal of radioactive waste and purification of radioactive contamination water. This paper aims to investigate the liquid phase dispersion inside co-current downward gas-liquid flow in the bed packed with spherical particle as basic model. Water and air are injected from the top center of column, and water is spreading to the entire cross-section flowing down along the column. Radial dispersion of water is captured by electrical resistance tomography on several axial positions. The results indicated that liquid flow rate have impact on radial dispersion of water.

The ability to predict void formation, void fraction and critical heat flux —CHF— in flow boiling under oscillatory flow and vibration conditions is important to the safety technology of nuclear reactor during earthquake. In the present study, the onset of nucleate boiling —ONB— and CHF on saturated flow boiling under vibration conditions were investigated experimentally. Steady state experiments were conducted using a copper thin-film and saturated and subcooled water at 0.1 MPa. The liquid velocity was 0.25, 1.38, 3.20 and 4.07 m/s, respectively; the liquid subcooling was 0 K and 20 K. A heater was made of a printed circuit board. A test section was a rectangular flow channel of 10 mm width and 10 mm height. The test heater was heated by Joule heating of d.c. current from a low-voltage high-current stabilizer. The heating rate of the heater was determined from supplied current and voltage. The temperature of the heater was obtained by referring to the measured electric resistance. The test section was arranged for horizontal position facing upward and for vertical position, respectively. For the vibration condition, the test section was set on a vibration table. The ONB was decided as an occurrence of the first boiling bubble. The critical heat flux was determined as that immediately before the heating surface physically burned-out. The CHF on saturated flow boiling under vibration conditions were investigated experimentally.