ICONE11-36099 CRITICAL POWER CHARACTERISTICS OF TIGHT LATTICE BUNDLES
スポンサーリンク
概要
著者
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Liu W
Japan Atomic Energy Agency
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Liu Wei
サウサンプトン大
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IWAMURA Takamichi
Japan Atomic Energy Research Institute
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Liu Wei
Japan Atomic Energy Agency
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Liu W
九大 大学院工学府
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Akimoto Hajime
Japan Atomic Energy Agency
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Akimoto H
Japan Atomic Energy Res. Inst. Ibaraki
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Akimoto Hajime
Frontier Research System For Global Change
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KURETA Masatoshi
Japan Atomic Energy Agency
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Iwamura Takamichi
Japan Atomic Energy Agency
関連論文
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- Effect of Rod Bowing on Critical Power Based on Tight-Lattice 37-Rod Bundle Experiments
- Gap Width Effect on Critical Power based on Tight-Lattice 37-Rod Bundle Experiments
- Pressure Drop Experiments using Tight-Lattice 37-Rod Bundles
- Critical Power Experiment with a Tight-Lattice 37-Rod Bundle
- Attempt on Detection of Natural Neutrinos by a Simple and Compact Apparatus
- Critical Power Characteristics in 37-rod Tight Lattice Bundles under Transient Conditions
- An Improved Critical Power Correlation for Tight-Lattice Rod Bundles
- Critical Power Correlation for Tight-Lattice Rod Bundles
- Critical Power in 7-Rod Tight Lattice Bundle(International Conferences on Power and Energy System)
- ICONE11-36099 CRITICAL POWER CHARACTERISTICS OF TIGHT LATTICE BUNDLES
- C′-4 用量増加実験における非線型効果の検討(日本統計学会第68回大会記録 : 医学統計 (2))
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- A MODELING STUDY OF LONG-RANGE TRANSPORT OF YELLOW SAND IN SPRING 2000
- Applicability of REFLA/TRAC Code to a Small-Break LOCA of PWR
- Pressure Drop Characteristics in Tight-Lattice Bundles for Reduced-Moderation Water Reactors(International Conferences on Power and Energy System)
- ICONE11-36172 EVALUATION OF CRITICAL HEAT FLUX OF TIGHT LATTICE CORE WITH SUBCHANNEL ANALYSIS CODE NASCA
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- PREDICTED THREE-DIMENSIONAL BUBBLY AND LIQUID FILM FLOW BEHAVIOR IN NARROW FUEL CHANNELS(Liquid Flow)
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- Application of PSA Methodology to Design Improvement of JAERI Passive Safety Reactor(JPSR)
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- Pressure Drop and Heat Transfer for Flow-Boiling of Water in Small-Diameter Tubes
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- Steam Water Pressure Drop under 15 MPa
- Development of Design Technology on Thermal-Hydraulic Performance in Tight-Lattice Rod Bundle: IV Large Paralleled Simulation by the Advanced Two-fluid Model Code
- Numerical Investigation of Cross Flow Phenomena in a Tight-Lattice Rod Bundle Using Advanced Interface Tracking Method
- Downcomer effective water head during reflood in postulated PWR LOCA.
- Assessment of current safety evaluation analysis on reflood behavior during PWR-LOCA by using CCTF data.
- Experimental study of system behavior during reflood phase of PWR-LOCA using CCTF.
- CHF Experiments under Steady-State and Transient Conditions for Tight Lattice Core with Non-Uniform Axial Power Distribution.
- Burnout characteristics under flow reduction condition.
- Oscillatory flows induced by direct contact condensation of flowing steam with injectes water.
- 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.
- Development of Design Technology on Thermal-Hydraulic Performance in Tight-Lattice Rod Bundles: I-Master Plan and Executive Summary
- Development of Design Technology on Thermal-Hydraulic Performance in Tight-Lattice Rod Bundles: II - Rod Bowing Effect on Boiling Transition under Transient Conditions
- Initial thermal-hydraulic behaviors under simultaneous ECC water injection into cold leg and upper plenum in a PWR-LOCA.
- Air-Water Two-Phase Cross Flow Resistance in Rod Bundle
- Experimental study of two-dimensional thermal-hydraulic behavior in core during reflood phase of 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.