Research contents

• 1 Development of the new thermal design method
  • 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 Heat flow experiment
  • 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 Application to the evaluation of heat flow in nuclear reactor systems
  • 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

1.3 Development of supercritical fluid heat transfer analysis method