The Multi-Fluid Multi-Phase Subchannel Analysis Code KAMUI for Subassembly Accident Analysis of an LMFR
スポンサーリンク
概要
- 論文の詳細を見る
A computer program KAMUI has been developed to simulate multi-dimensional thermal-hydraulic behaviors of the coolant and disrupted fuels in a subassembly under hypothetical accident conditions of a Liquid-Metal Fast Reactor (LMFR) where the strong momentum and thermal coupling of the coolant flow dominates the material relocation. The KAMUI code is based on the subchannel analysis approach to model the fuel pin array geometry and has a two-velocity field, multi-phase/multi-component formulation for transport phenomena with relatively simple but conventional constitutive models. The code has been applied to and validated for a total inlet blockage simulation experiment SCARABEE BE+1. The results show the significance of the radial heat loss through the wrapper tube wall and its influences on the boiling evolution and dynamic behaviors of the coolant and molten clad motion. They also indicate that the multi-dimensional analysis is essential in evaluating hypothetical subassembly accidents of LMFRs including a large-scale local blockage and a total inlet blockage.
- 一般社団法人 日本原子力学会の論文
- 2000-08-25
著者
-
NINOKATA Hisashi
Research Laboratory for Nuclear Reactors, Tokyo Institute of Technology
-
Ninokata Hisashi
Research Lab. For Nuclear Reactors Tokyo Institute Of Technology
-
Kasahara Fumio
Nuclear Power Engineering Corporation
-
KASAHARA Fumio
Research Laboratory for Nuclear Reactors, Tokyo Institute of Technology
関連論文
- Study of the Self-Controllability for the Fast Reactor Core with High-Thermal-Conductivity Fuel
- Numerical Studies on Dynamic Behavior of Air-Water Cross Flow Between Two Circular Interconnected Channels
- Study of the Self-Controllability for the Fast Reactor Core with High-Thermal-Conductivity Fuel
- Computational Fluid Dynamics Study of Liquid Droplet Impingement Erosion in the Inner Wall of a Bent Pipe
- PRA-Based SMA : the First Tool toward a Risk-Informed Approach to the Seismic Design of the IRIS
- Analysis on Pipe Rupture of Steam Condensation Line at Hamaoka-1, (II) Hydrogen Combustion and Pipe Deformation
- Analysis on Pipe Rupture of Steam Condensing Line at Hamaoka-1, (I) Accumulation of Non-condensable Gas in a Pipe
- Prediction of Release Rate of Burnt Sodium as Aerosol
- ICONE11-36048 Development of Boiling Transition Analysis Code TCAPE-INS/B based on Mechanistic Methods for BWR Fuel Bundles : Models and verification with the boiling transition experimental data
- Analysis of an Out-of-pile Experiment for Materials Redistribution under Core Disruptive Accident Condition of Fast Breeder Reactors
- Prediction of the Equilibrium Two-Phase Flow Distributions in Inter-Connected Subchannel Systems
- Concept of Erbium Doped Uranium Oxide Fuel Cycle in Light Water Reactors
- Second International Symposium on Global Environment and Nuclear Energy Systems
- Numerical Method for Simulation of Fluid Flow and Heat Transfer in Geometrically Disturbed Rod Bundles
- Numerical Study on Observed Decay Ratio of Coupled Neutronic-Thermal Hydraulic Instability in Ringhals Unit 1 under Random Noise Excitation
- An Algorithm for Attenuation of Turbulence in Particulate Flow Linked to the Fluid-dynamic Code COMMIX-M
- Development of Parallel Coupling System between Three-Dimensional Nodal Kinetic Code ENTREE and Two-Fluid Plant Simulator TRAC/BF1
- The Multi-Fluid Multi-Phase Subchannel Analysis Code KAMUI for Subassembly Accident Analysis of an LMFR
- Nonlinear Iterative Nodal Method Applied to Neutron Flux Modal Analysis
- Comparison of sodium evaporations measured under condition of fog formation.