Experimental study of upper core quench in PWR reflood phase.
スポンサーリンク
概要
- 論文の詳細を見る
The present study investigates experimentally the characteristics of quench phenomena, top quench and bottom quench, in the upper part of the core during reflood phase in a PWR-LOCA. The characteristics of quench phenomena in the upper part of the core have not been predicted well even by a best estimate safety analysis code such as TRAC. The following are revealed from the experiment. (1) Quench phenomena in the upper part of the core are classified into three types ; (1) top quench of heated rods, (2) top quench of non-heated rods, and (3) bottom quench of heated rods, with respect to the relation of quench velocity vs. quench temperature. (2) Top quench of heated rods is influenced by the falling water film on the adjacent non-heated rod. (3) Top quench velocity of heated rod rises with an increase of quench temperature ; this is a tendency opposite to the prediction by the existing theory. (4) Top quench of non-heated rods has the same tendency as the existing theory.<BR>The effects of system pressure and coolant flow rate are also investigated.
- 一般社団法人 日本原子力学会の論文
著者
-
Sudo Yukio
Japan Atomic Energy Research Institute
-
OSAKABE Masahiro
Japan Atomic Energy Research Institute
-
ABE Yutaka
Japan Atomic Energy Research Institute
関連論文
- 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
- 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.
- Simulation of PWR Turbine Trip Transient in ROSA-IV Large Scale Test Facility
- 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.
- Investigation of break orientation effect during cold leg small-break LOCA at ROSA-IV LSTF.
- 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
- Film boiling heat transfer during reflood phase in postulated PWR loss-of-coolant accident.
- Elimination of Numerical Pressure Spikes Induced by Two-Fluid Model.