Experimental Study of Falling Water Limitation under a Counter-Current Flow in a Vertical Rectangular Channel : 1st Report, Effect of Flow Channel Configuration and Introduction of CCFL Correlation
スポンサーリンク
概要
- 論文の詳細を見る
Counter-current-flow limitation (CCFL) experiments were carried out for both vertical rectangular channels and vertical circular tubes varying in size and in configuration of their cross sections to clarify CCFL characteristics in the vertical rectangular channels. Quantitative understanding of critical heat flux (CHF) in a narrow vertical rectangular channel is required for the thermohydraulic design and safety analysis of research nuclear reactors in which flat-plate-type fuel is employed. Critical heat flux under downward low velocity is closely related to falling water limitation under a counter-current flow. Experimental results showed that the equivalent hydraulic diameter de, i.e., the width, and gap of the channel play an important role in determining the CCFL characteristics of a rectangular channel. However, a significant influence of channel length on CCFL characteristics was not observed in the ranges investigated. Using new dimensionless parameters, the authors propose a correlation for predicting the relationship between upward air velocity and downward water velocity based on the present experimental results.
- 一般社団法人日本機械学会の論文
- 1991-05-15
著者
-
Kaminaga Masanori
Japan Atomic Energy Research Institute
-
Usui Tohru
Nkk Corporation
-
Sudo Yukio
Japan Atomic Energy Research Institute
関連論文
- ICONE11-36128 THERMAL-HYDRAULIC EXPERIMENTS AND ANALYSES FOR COLD MODERATORES
- Experimental Study of Falling Water Limitation under a Counter-Current Flow in a Vertical Rectangular Channel : 1st Report, Effect of Flow Channel Configuration and Introduction of CCFL Correlation
- Heat Transfer Characteristics in Narrow Vertical Rectangular Channels Heated from Both Sides
- Experiments on Mercury Circulation System for Spallation Neutron Target
- ICONE11-36146 WATER FLOW EXPERIMENTS AND ANALYSES ON THE CROSS-FLOW TYPE MERCURY TARGET WITH PERFORATED PLATES
- ICONE11-36101 MERCURY TARGET THERMAL HYDRAULIC DEDIGN FOR JAERI SPALLATION NEUTRON SOURCE
- ICONE11-36079 Mercury Erosion Experiments for Spallation Target System
- ICONE11-36094 STRUCTURAL INTEGRITY OF BEAM WINDOW OF MERCURY TARGET
- Improvement of Critical Heat Flux Correlation for Research Reactors using Plate-Type Fuel
- Mechanism of Falling Water Limitation under Counter-current Flow through a Vertical Flow Path
- Experimental Study of Homogeneously Dispersed Two-phase Critical Flow
- Core thermohydraulic design with 20% LEU fuel for upgraded research reactor JRR-3.
- Experimental study of incipient nucleate boiling in narrow vertical rectangular channel simulating subchannel of upgraded JRR-3.
- Analytical study on thermal-hydraulic behavior of transient from forced circulation to natural circulation in JRR-3.
- Experimental study of differences in DNB heat flux between upflow and downflow in vertical rectangular channel.
- Heat Transfer Calculation of Simulated Heater Rods throughout Reflood Phase in Postulated PWR-LOCA Experiments
- Estimation of average void fraction in vertical two-phase flow channel under low liquid velocity.
- Evaluation of Local Power Distribution with Fine-mesh Core Model for High Temperature Engineering Test Reactor (HTTR).
- Effects of partial flow blockage on core heat transfer in forced-feed reflood tests.
- Upper plenum dump during reflood in PWR loss-of-coolant accident.
- Prediction of Dryout Heat Flux for Particle Bed Simulating Degraded Core in LWR Severe Core Damage Accidents
- Combined forced and free convective heat transfer characteristics in narrow vertical rectangular channel heated from both sides.
- Downcomer effective water head during reflood in postulated PWR LOCA.
- Experimental study of differences in single-phase forced-convection heat transfer characteristics between upflow and downflow for narrow rectangular channel.
- Experimental study of upper core quench in PWR reflood phase.
- Parameter effects on downcomer penetration of ECC water in PWR-LOCA.
- Analysis of saturated film boiling heat transfer in reflood phase of PWR-LOCA. Turbulent boundary layer model.:Turbulent Boundary Layer Model
- Film boiling heat transfer during reflood phase in postulated PWR loss-of-coolant accident.