The most common approach to classify the flow, which can be found in various reactors, is to divide into 'single phase flow' and 'multi phase flow'. Water flow without boiling, or air flow is typical example of single phase flow. In thermal design of nuclear reactor, it is often the case that 'gas-liquid two phase flow' or single phase flow is treated. Therefore, the focus of our study is on development of the simulation code, which can treat these two kinds of flow.
1.1 Development of two-phase flow analysis using the phase-field interface-tracking method
1.1.1 Liquid film flow, bubble flow, spray flow
1.1.2 Phase transition model
1.1.3 Bubble flow behavior in the nuclear reactor during an earthquake
1.2 Development of two-phase flow analysis in the nuclear reactor using a three-dimensional two-fluid model
1.2.1 Bundle system analysis
1.2.2 Coupled thermo-mechanical finite-element analysis
1.3 Development of supercritical fluid heat transfer analysis method
1.4 Development of multiphase flow analysis
1.5 Development of systems analysis techniques in the nuclear reactor
2.1 Development of heat flow measurement
2.1.1 Technique for coincident measurement of surface temperature and heat flux
2.1.2 Measurement technique for void fraction of capacitance type
2.1.3 Measurement technique for void fraction distribution by using wire meshes
2.1.4 Measurement technique for droplet volume
2.2 Model experiments
2.2.1 Experiment for boiling heat transfer
2.2.2 Experiment for cross-flow of two parallel flows
3.1 Characterization of heat transfer in the heat transfer tube of FBR
3.1.1 Experiment for two-phase flow in the heat transfer tube of FBR
3.2 Thermal Feasibility of water reactor for Flexible Fuel Cycle (FLWR)
3.2.1 37-rod bundles test
3.2.2 Experimantal estimation for the Effect of Axial Power Distribution on Critical Power
3.3 Thermal characterization supercritical pressure light water reactor
3.3.1 19-rod bundles analysis
3.4 Severe phenomenon evaluation
3.4.1 Analysis for phenomena of Fukushima accident
3.4.2 Deformation and phase transition of the structure materials in the nuclear reactor
3.4.3 Analysis of migration behavior of melting debris in the nuclear reactor
3.4.4 Jet break behavior of melting materials
3.4.5 Analysis of detailed temperature distribution in the nuclear reactor
3.4.6 Experiment for heat flow in the debris
3.4.7 Experiment for heat transfer of seawater
3.4.8 Test measurement of void fraction distribution in fuel assembly