Current Project

 

Development of a Computational Technique Assessing Core Cooling Capability during Severe Accidents

  • Evaluating the core cooling capability of OPR1000 and APR1400 using the MELCOR code, which is one of the best-optimized computational tools for analysis of hypothetical severe accidents
  • Contributing on
    • Establishing an effective assessment methodology as well as technique of core cooling capability
    • Constructing a core cooling capability map, which will facilitate archive a severe accident management guideline against the hypothesized severe accident event

 

research1

 

 

Development of Inherent Indicator for Thermal Flow Measurement of Liquid Sodium

Improvement and domestic production of the indicators measuring inherent thermal flow characteristics of liquid sodium

  •  Evaluation of performance and compatibility of the sensor and indicator imported from IPPE, a Russian sensor engineering company
  • Development of technical methodologies for improving performance of the IPPE’s indicator and prototypical design for its domestic production
  • Manufacturing and performance test of the prototypical indicator

 

 

Advanced Thermal-Hydraulic Experiments for Nuclear Applications Laboratory

　1. Integrated thermal-hydraulics laboratory development to enhance engineering school education

  • Building advanced nuclear thermal-hydraulics experiment facility and auxiliary measurement system
  • Developing and operating on the experimental program targeted to undergraduate student
  • Building a homepage for integrated experimental education including experimental database

 

research3-1

　2. Forced convective heat transfer experiments

  • Single-phase flow heat transfer experiment
    • Measuring for heat transfer coefficient and pressure drop through annular channel on condition of laminar and turbulent flow
    • Observing a developing process of single-phase convection through visualization channel
  • Onset of Nucleate Boiling (ONB) measurement
    • Visualization of ONB phenomenon in narrow rectangular channel on condition of atmosphere pressure
    • Collecting the information of ONB occurrence at core condition of research reactor
    • Using sputtering technique to make surface condition same with research reactor
  • Two-phase heat transfer experiment
    • Measuring Nucleate Boiling Heat Transfer Coefficient (NBHTC) and Critical Heat Flux (CHF) based on flow condition inside core of PWR realized by using fluid-to-fluid modeling technique
    • Enhancing phenomenon understanding for physical mechanism of various flow regime through flow visualization in annular channel
    • Evaluating improvement of heat transfer performance of fuel cladding using sputtering technique to change surface condition

 

Boiling_Curve