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

 

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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

 

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 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