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3 Application to the evaluation of heat flow in nuclear reactor systems

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.


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