Interfacial drag coefficient of air-water mixture in rod bundle.
スポンサーリンク
概要
著者
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TASAKA Kanji
Japan Atomic Energy Research Institute
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KOIZUMI Yasuo
Japan Atomic Energy Research Institute
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OSAKABE Masahiro
Japan Atomic Energy Research Institute
関連論文
- Simulation of PWR Turbine Trip Transient in ROSA-IV Large Scale Test Facility
- Heat Transfer Calculation of Simulated Heater Rods throughout Reflood Phase in Postulated PWR-LOCA Experiments
- Simulation Test of PWR Instrument Tube Break LOCA in ROSA-IV Program
- Effects of partial flow blockage on core heat transfer in forced-feed reflood tests.
- Experimental study of upper core quench in PWR reflood phase.
- Investigation of break orientation effect during cold leg small-break LOCA at ROSA-IV LSTF.
- Simulation experiment of five percent small break loss-of-coolant accident of boiling water reactor.
- Interfacial drag coefficient of air-water mixture in rod bundle.
- Characteristic of two-phase slanting flow in rod bundle.
- Two-phase flow pattern and heat transfer during core uncovery.
- Analysis of saturated film boiling heat transfer in reflood phase of PWR-LOCA. Turbulent boundary layer model.:Turbulent Boundary Layer Model
- Investigation of pre- and post-dryout heat transfer in upward steam-water two-phase flow at low flow rate.
- Simplified method for calculation of neutron capture transformation effects of fission products on decay power.
- Build-up and decay of actinide nuclides in fuel cycle of nuclear reactors.
- Core liquid level depression due to manometric effect during PWR small break LOCA. Effect of break area.:Effect of Break Area