Kazuyuki Takase

Direct numerical simulation of turbulent channel flow under a uniform magnetic field for large-scale structures at high Reynolds number

A direct numerical simulation (DNS) of turbulent channel flow with high Reynolds number has been carried out to show the effects of the magnetic field. In this study, the Reynolds number for channel flow based on bulk velocity Ub, viscosity ν, and channel width 2δ was set to be constant; Reb=2δUb∕ν=45818. A uniform magnetic field was applied in the direction of the wall normal. The value of the Hartmann number, Ha were 32.5 and 65, where Ha=2δB0σ∕ρν. The turbulent quantities such as the mean flow, turbulent stress, and turbulent statistics were obtained by DNS. Although the influence of the magnetohydrodynamic dissipation terms in the turbulent kinetic energy budget was small, large-scale turbulent structures, e.g., vertical structures, low-speed streaks, ejection, and sweep, were found to decrease at the central region of the channel. Consequently, the difference between production and dissipation in the turbulent kinetic energy decreased with increasing Hartmann number at the central region and large-sc...

Experimental Study on Buoyancy-Driven Exchange Flows Through Breaches of a Tokamak Vacuum Vessel in a Fusion Reactor Under the Loss-of-Vacuum-Event Condition

As one of thermofluid safety studies in the International Thermonuclear Experimental Reactor, buoyancy-driven exchange flow behavior through breaches of a vacuum vessel (VV) has been investigated quantitatively by using a preliminary low-of-vacuum-event (LOVA) apparatus that simulated the tokamak VV of a fusion reactor with a small-scaled model. To carry out the present experiments under the atmospheric pressure condition, helium gas and air were provided as the working fluids. The inside of the VV was initially filled with helium gas and the outside was atmosphere. The breaches on the VV under the LO VA condition were simulated by opening six simulated breaches to which were set the different positions on the VV. When the buoyancy-driven exchange flow through the breach occurred, helium gas went out from the inside of the VV through the breach to the outside and air flowed into the inside of the VV through the breach from the outside. The exchange rate in the VV between helium gas and air was calculated from the measured weight change of the VV with time since the experiment has started. Experimental parameters were breach position, breach number, breach length, breach size, and breach combination. The present study clarifies that the relation between the exchange rate and the breach position of the VV depended on the magnitude of the potential energy from the ground level to the breach position, and then, the exchange rate decreased as the breach length increased and as the breach size decreased.

Development of Analytical Procedures of Two-Phase Flow in Tight-Lattice Fuel Bundles for Innovative Water Reactor for Flexible Fuel Cycle

Abstract A research and development project to investigate thermal-hydraulic performance in the tight-lattice rod bundles of the Innovative Water Reactor for Flexible Fuel Cycle (FLWR) has been in progress at Japan Atomic Energy Agency in collaboration with power companies, reactor vendors, and universities since 2002. The FLWR can realize favorable characteristics such as effective utilization of uranium resources, multiple recycling of plutonium, high burnup, and long operation cycle, based on matured light water reactor technologies. Mixed-oxide fuel assemblies with tight lattice arrangement are used because they increase the conversion ratio by reducing the moderation of neutrons. Increasing the in-core void fraction also contributes to the reduction of neutron moderation. Information about the effects of the gap width and grid spacer configuration on the flow characteristics in the FLWR core is still insufficient. Thus, we are developing procedures for qualitative analysis of thermal-hydraulic performance of the FLWR core using an advanced numerical simulation technology. In this study, an advanced two-fluid model is developed to economize on the computing resources. In the model, interface structures larger than computational cells (such as liquid film) are simulated by the interface tracking method, and small bubbles and droplets are estimated by the two-fluid model. In this paper, we describe the outline of this model and the numerical simulations we performed to validate the model performance qualitatively.

Results of Two-Phase Flow Experiments with an Integrated Ingress-of-Coolant Event (ICE) Test Facility for ITER Safety

Abstract An integrated ICE (Ingress-of-Coolant Event) test facility was constructed in order to demonstrate the adequacy of the ITER (International Thermonuclear Experimental Reactor) safety design approach and investigate two-phase flow behavior during an ICE event. The integrated ICE test facility corresponds to approximately 1/1600 of the ITER components. In the experiments the fluid flow configurations inside the vacuum vessel at the ICE events were observed visually and pressure transients were measured quantitatively. Two-phase flow analyses were carried out with the TRAC code and the experimental results were validated numerically. From the present study it was clarified that the ITER pressure suppression system is very effective in reducing the pressure rise in case of the ICE event and the water–vapor two-phase flow characteristics in ITER can be predicted numerically with sufficient accuracy.

Development of Numerical Simulation Method for Relocation Behavior of Molten Materials in Nuclear Reactors: Analysis of Relocation Behavior for Molten Materials With a Simulated Decay Heat Model

In order to precisely investigate molten core relocation behavior in the Fukushima Daiichi nuclear power plants, we have developed the detailed and phenomenological numerical simulation code named JUPITER for predicting the molten core behavior including solidification and relocation based on the three-dimensional multiphase thermal-hydraulic simulation models. In this paper, in order to distinguish a fuel component from core internals component in the JUPITER code, we added the multicomponent analysis method to the code and carried out to check effectiveness of the multicomponent analysis model based on the numerical simulation of melting behavior of the simulated fuel assemblies and core internals. From the present numerical results, it was confirmed that a newly developed multicomponent analysis method appropriately can predict the relocation behavior of molten materials in complicated structures, that is relocation and solidification behavior for fuel components including the heat source, i.e., simulated decay heat, and the core internals without heat sources from simplified fuel assemblies through the core support plate and simplified control rod guide tubes to the lower head in the reactor pressure vessel (RPV).Copyright

Development of Prediction Technology of Two-Phase Flow Dynamics Under Earthquake Acceleration — (4) Measurement of Bubble Deformation Near Wall Under Flow Rate Fluctuation

In order to clear the two-phase flow behavior under earthquake, a systematic study is done experimentally and numerically. The present study is one on the series of the study on two-phase flow under earthquake, and focuses on the flow rate fluctuation. The flow rate fluctuation was added to bubbly or plug flow in a horizontal pipe, and flow behavior was measured by PIV and image processing. The bubble deformation near the pipe wall was observed and the velocity field around the bubble was shown. The bubble coalescence was also observed under the flow rate fluctuation condition.Copyright

Measurement of Surface Heat Flux and Surface Temperature in Nucleate Pool Boiling Using Micro-Thermocouples

In this paper, a measurement system for surface temperature and surface heat flux was developed to study heat transfer mechanism in boiling process. The system was consisted by two parts: (1) inner block temperatures were measured using micro-thermocouples set at two layers inside heating block; (2) with using the measured temperatures, inverse heat transfer analysis was performed to get surface heat flux and surface temperature. For the inner block temperature measurement, special T-type micro thermocouples with a common positive pole were developed. Totally 20 thermocouples were set at two layers at the depths 3.1μm and 4.905mm beneath the boiling surface, in a radius of 5mm. The developed system was used to research the change of surface heat flux and surface temperature in a boiling process. Experiments were performed to pool boiling at atmospheric pressure. The experiments showed the developed special T-type micro thermocouples could trace temperature change in boiling process successfully. With comparison to images from a high-speed camera, temperature change tendencies in boiling process were tried to understand. Then one dimensional inverse heat conduction problem was solved to get surface heat flux and surface temperature. Increase in surface heat flux with the generation of big bubble was derived successfully.Copyright

Thermal Feasibility Analyses for the 1356MWe High Conversion-Type Innovative Water Reactor for Flexible Fuel Cycle

A new reactor concept of innovative water reactor for flexible fuel cycle (FLWR) is under development at Japan Atomic Energy Agency in cooperation with Japanese reactor suppliers. A design of 1,356 MWe high conversion boiling water reactor-type FLWR core, which has an instantaneous conversion ratio of 1.04, negative void coefficient, high burnup of 65 GWd/t, and 15-month operational cycle length, has been constructed. So far, studies on thermal-hydraulic characteristics have been performed for tight lattice core. Evaluation methods for the critical power and the pressure drop under both the steady and the transient states have been established, and a modified TRAC-BF1 code has been developed for the thermal-hydraulic design of the FLWR. In this paper, the thermal feasibility of the designed 1356MWe FLWR core is analyzed by using the modified TRAC-BF1 code. The analysis is first carried out for the current core design. It is confirmed that no boiling transition (BT) occurs under the steady state. However, the minimum critical power ratio (MCPR) is only about 1.08, and the BT is confirmed occurring under the postulated abnormal transient processes. Therefore, concretizations of the conditions that ensure the thermal feasibility of a natural circulation-type FLWR and a forced circulation-type FLWR are performed. As for the results, for a forced circulation-type FLWR, the operation-limited MCPR (OLMCPR) is 1.32, and the necessary minimum core coolant flow rate is 640 kg/(m2s). For a natural circulation-type FLWR, the OLMCPR is 1.19, and the necessary minimum core coolant flow rate is 560 kg/(m2s).

Pressure Propagation and Critical Flow (Compressible Fluid Flow)

In the previous chapters, discussion was devoted mainly to a fluid for which there is no need to consider the fluid compressibility. In this chapter, we consider flows in a wide range of temperatures and pressures and treat each gas as a compressible fluid. In such a case, the sound velocity plays a unique role, and the Mach number of the flow becomes an important parameter that determines the flow characteristics. In addition to air, a wide variety of gases, including steam and combustion gases, are considered here. Most of these gases are not ideal gases; however, since they may be regarded as approximating ideal gases due to the high temperature, the computational formulas for ideal gases are described, unless otherwise noted. This treatment is an approximation; however, it is sufficient to present qualitative features and provide theoretical results with good reliability in many cases. In addition, the flow is assumed to be steady. Another treatment must be considered when an unsteady flow has an important role; however, since that leads to a complicated analysis and significantly limits the generality of the conditions considered, no explanation is given here.

Development of Prediction Technology of Two-Phase Flow Dynamics Under Earthquake Acceleration: (13) Rising Bubble Motion Under Horizontal Vibration

An earthquake is one of the most serious phenomena for the safety of a nuclear reactor in Japan. Therefore, structural safety of nuclear reactors has been studied and nuclear reactors ware contracted with structural safety for a big earthquake. However, it is not enough for safety operation of nuclear reactors because thermal-fluid safety is not confirmed under the earthquake. For instance, behavior of gas-liquid two-phase flow is unknown under the earthquake conditions. Especially, fluctuation of void fraction is an important factor for the safety operation of the nuclear reactor. In the previous work, fluctuation of void faction in bubbly flow was studied experimentally and theoretically, to investigate the stability of the bubbly flow. In such studies, flow rate or void fraction fluctuations were given to the steady bubbly flow. In the case of the earthquake, the fluctuation is not only the flow rate, but also a body force on the two-phase flow and a shear force through a pipe wall. Interactions of gas and liquid through their interface also act on the behavior of the two-phase flow. The fluctuation of the void fraction is not clear for such complicated situation under the earthquake.Therefore, in this research project, the behavior of gas-liquid two-phase flow is investigated experimentally and numerically in the series of study. In this study, to investigate the effects of vibration on bubbly flow in the components and construct an experimental database for validation, we performed visualization experiments of vertical bubbly flow in a rectangular water tank on which a sine wave vibration was applied. In this paper, results of visualized experiment evaluated by the visualization techniques, including positions of bubbles, shapes of bubbles and liquid velocity distributions around bubbles, were shown. And liquid velocity distribution around bubbles by the PIV measurement was also shown. In the results, bubble behaviors were affected by oscillation. And the cycle of the bubble tilt angle was almost same as the cycle of oscillation table velocity.Copyright