MURAO Yoshio | Japan Atomic Energy Research Institute

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概要

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著作論文

A Concept of Passive Safety Pressurized Water Reactor System with Inherent Matching Nature of Core Heat Generation and Heat Removal
Applicability of REFLA/TRAC Code to a Small-Break LOCA of PWR
Application of PSA Methodology to Design Improvement of JAERI Passive Safety Reactor(JPSR)
Transient Analysis for Design of Primary Coolant Pump Adopted to JAERI Passive Safety Reactor JPSR
Possibility of a Pressurized Water Reactor Concept with Highly Inherent Heat Removal Following Capability
Evaluation on Driving Force of Natural Circulation in Downcomer for Passive Residual Heat Removal System in JAERI Passive Safety Reactor JPSR
Assessment of current safety evaluation analysis on reflood behavior during PWR-LOCA by using CCTF data.
Experimental study of ECC water injection rate effects on reflood phase of PWR-LOCA.
Experimental study of initial downcomer water accumulation velocity effects on reflood phase of PWR-LOCA.
Assessment of core thermo-hydrodynamic models of REFLA-1D code with CCTF data for reflood phase of PWR-LOCA.
Experimental study of system behavior during reflood phase of PWR-LOCA using CCTF.
Effect of loop seal on reflood phenomena in PWR.
CHF Experiments under Steady-State and Transient Conditions for Tight Lattice Core with Non-Uniform Axial Power Distribution.
Experimental Study on Difference in Reflood Core Heat Transfer among CCTF, FLECHT-SET and Predicted with FLECHT Correlation.
Core radial power profile effect on system and core cooling behavior during reflood phase of PWR-LOCA with CCTF data.
System pressure effect on system and core cooling behavior during reflood phase of PWR LOCA.
Water accumulation phenomena in upper plenum during reflood phase of PWR-LOCA by using CCTF data.
Experimental study on reflood behavior in PWR with upper plenum injection type ECCS by using CCTF.
Visual study of flow behavior in upper plenum during simulated reflood phase of PWR-LOCA.
Effect of decay heat level on reflood phenomena during PWR-LOCA.
Assessment of core radial power profile effect model for REFLA code by using CCTF data.
Pressure drop through broken cold leg during reflood phase of loss-of-coolant accident of pressurized water reactor.
Applicability of Core Thermal-Hydraulic Models in REFLA Code to 17*17 Type Fuel Assembly of PWR.
Elimination of Numerical Pressure Spikes Induced by Two-Fluid Model.

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