IGUCHI Tadashi | Japan Atomic Energy Research Institute
スポンサーリンク
概要
関連著者
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IGUCHI Tadashi
Japan Atomic Energy Research Institute
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MURAO Yoshio
Japan Atomic Energy Research Institute
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Iguchi Tadashi
Japan Atomic Energy Research Institute Department Of Reactor Safety Research
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OKUBO Tsutomu
Japan Atomic Energy Agency
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Akimoto Hajime
Japan Atomic Energy Agency
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NAKAMURA Hideo
Japan Atomic Energy Agency
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SIBAMOTO Yasuteru
Japan Atomic Energy Agency
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IGUCHI Tadashi
Japan Atomic Energy Research Institute, Department of Reactor Safety Research
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KUKITA Yutaka
Department of Energy Engineering and Science, Nagoya University
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Kukita Yutaka
Department Of Energy Engineering And Science Nagoya University
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Okubo Tsutomu
Japan Atomic Energy Research Institute
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Kukita Y
Department Of Energy Engineering And Science Nagoya University
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Kukita Yutaka
Nagoya University
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Asaka Hideo
JAPAN Atomic Energy Research Institute
著作論文
- ICONE11-36298 SYSTEM PRESSURE EFFECT ON DENSITTY-WAVE INSTABILITY : SIMPLIFIED MODEL ANALYSIS AND EXPERIMENTS
- ICONE11-36452 EXPERIMENTAL STUDY ON COOLING LIMIT UNDER FLOW INSTABILITY IN BOILING FLOW CHANNEL
- Experimental study of initial downcomer water accumulation velocity effects on reflood phase of PWR-LOCA.
- Effect of loop seal on reflood phenomena in PWR.
- 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.
- Void fraction in simulated PWR fuel bundle during reflood phase.
- 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.