Elimination of Numerical Pressure Spikes Induced by Two-Fluid Model.
スポンサーリンク
概要
- 論文の詳細を見る
In the initial stage of reflood phase of PWR-LOCA, quasi-stable surface of water is established in a core. Spurious pressure spikes are often encountered when two-fluid model codes are used to analyze two-phase flow dynamics in the initial stage of reflood phase. These pressure spikes are not observed in experiments. Since these pressure spikes affect other variables such as void fraction, it is important to eliminate these pressure spikes to get physically reasonable results with two-fluid model codes. In the present study, it is quantitatively clarified that these pressure spikes result from numerical acceleration loss of liquid above the surface of water where liquid does not exist. Furthermore, a method is developed to mitigate the acceleration loss of liquid above the surface of water. It is confirmed that this method is effective to eliminate these pressure spikes without losing benefits of the present two-fluid model codes.
- 一般社団法人 日本原子力学会の論文
著者
-
Akimoto Hajime
Japan Atomic Energy Agency
-
MURAO Yoshio
Japan Atomic Energy Research Institute
-
ABE Yutaka
Japan Atomic Energy Research Institute
-
KAMO Hideki
Japan Research Institute, Ltd.
関連論文
- 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
- 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
- A Concept of Passive Safety Pressurized Water Reactor System with Inherent Matching Nature of Core Heat Generation and Heat Removal
- 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
- ICONE11-36098 PRESSURE DROP CHARACTERISTICS IN TIGHT-LATTICE ROD BUNDLES FOR REDUCED-MODERATION WATER REACTERS
- PREDICTED THREE-DIMENSIONAL BUBBLY AND LIQUID FILM FLOW BEHAVIOR IN NARROW FUEL CHANNELS(Liquid Flow)
- ICONE11-36097 DEVELOPMENT OF MECHANISTIC BOILING TRANSITION MODEL IN ROD BUNDLES
- Application of PSA Methodology to Design Improvement of JAERI Passive Safety Reactor(JPSR)
- Transient Analysis for Design of Primary Coolant Pump Adopted to JAERI Passive Safety Reactor JPSR
- Possibility of a Pressurized Water Reactor Concept with Highly Inherent Heat Removal Following Capability
- Evaluation on Driving Force of Natural Circulation in Downcomer for Passive Residual Heat Removal System in JAERI Passive Safety Reactor JPSR
- FLOW VISUALIZATION OF A WAKE STRUCTURE AROUND THIN PLATE BY USING DYNAMIC PIV SYSTEM(Measurement)
- Model Development for Bubble Turbulent Diffusion and Bubble Diameter in Large Vertical Pipes
- Numerical Investigation of Heat Transfer Enhancement Phenomenon during the Reflood Phase of PWR-LOCA
- Development of Design Technology on Thermal-Hydraulic Performance in Tight-Lattice Rod Bundles: III - Numerical Evaluation of Fluid Mixing Phenomena using Advanced Interface-Tracking Method -
- Development of Design Technology on Thermal-hydraulic Performance in Tight-lattice Rod Bundles: V-Estimation of Void Fraction
- 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
- Prediction of Dryout Heat Flux for Particle Bed Simulating Degraded Core in LWR Severe Core Damage Accidents
- Downcomer effective water head during reflood in postulated PWR LOCA.
- Experimental study of upper core quench in PWR reflood phase.
- Assessment of current safety evaluation analysis on reflood behavior during PWR-LOCA by using CCTF data.
- Experimental study of ECC water injection rate effects on reflood phase of PWR-LOCA.
- Experimental study of initial downcomer water accumulation velocity effects on reflood phase of PWR-LOCA.
- Assessment of core thermo-hydrodynamic models of REFLA-1D code with CCTF data for reflood phase of PWR-LOCA.
- Experimental study of system behavior during reflood phase of PWR-LOCA using CCTF.
- Effect of loop seal on reflood phenomena in PWR.
- CHF Experiments under Steady-State and Transient Conditions for Tight Lattice Core with Non-Uniform Axial Power Distribution.
- Experimental Study on Difference in Reflood Core Heat Transfer among CCTF, FLECHT-SET and Predicted with FLECHT Correlation.
- 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
- 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.
- 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.