Study of the Self-Controllability for the Fast Reactor Core with High-Thermal-Conductivity Fuel

スポンサーリンク

概要

• 論文の詳細を見る

• The tolerance capability against ATWS for the FBR core with metallic fuel can be improved by employing a fuel with high thermal conductivity (HTC fuel) instead of the conventional metallic fuel, U-Pu-Zr. To investigate the self-controllability for the HTC-fueled core with U-Pu-Al alloy fuel, having one order of magnitude higher thermal conductivity than that of the U-Pu-Zr, the core employing the U-Pu-Al fuel was evaluated against ULOF and UTOP. Based on the systematic calculation, it was found that the larger temperature margin between the steady state and ULOF/UTOP conditions caused the excellent tolerance capability against ULOF and UTOP for the HTC-fueled core compared with that for the Zr-alloy-fueled core. Also, the conditions of the core reactivity coefficients required for neither fuel melting nor coolant boiling were investigated by using a “self-controllability map” consisting of effective fuel and coolant reactivities. As a result, the self-controllable region was found to be expanded especially for UTOP in the case of the HTC-fueled core.

• 2010-08-01

著者

• Ishizu Tomoko

  Research Laboratory for Nuclear Reactors, Tokyo Institute of Technology

• Yokoyama Tsugio

  Research Laboratory for Nuclear Reactors, Tokyo Institute of Technology

• Endo Hiroshi

  Nuclear Energy System Safety Div., Japan Nuclear Energy Safety Organization (JNES)

• TOKIWAI Moriyasu

  Komae Lab., Central Research Institute of Electric Power Industry (CRIEPI)

• NINOKATA Hisashi

  Research Laboratory for Nuclear Reactors, Tokyo Institute of Technology

• NINOKATA Hisashi

  Department of Nuclear Engineering, Tokyo Institute of Technology

• Ishizu Tomoko

  Research Laboratory For Nuclear Reactors Tokyo Institute Of Technology

• Ishizu Tomoko

  Department Of Nuclear Engineering Tokyo Institute Of Technology

• Ninokata Hisashi

  Tokyo Inst. Of Technol.

• Ninokata Hisashi

  Res. Lab. For Nuclear Reactors Tokyo Inst. Of Technol.

• Ninokata Hisashi

  Research Lab. For Nuclear Reactors Tokyo Institute Of Technology

• Tokiwai Moriyasu

  Komae Lab. Central Research Institute Of Electric Power Industry (criepi)

• Yokoyama Tsugio

  Research Laboratory For Nuclear Reactors Tokyo Institute Of Technology

• Endo Hiroshi

  Nuclear Energy System Safety Div. Japan Nuclear Energy Safety Organization (jnes)

関連論文

• Study of the Self-Controllability for the Fast Reactor Core with High-Thermal-Conductivity Fuel
• On Void Reactivity Limitation for the Core Loaded with Mixed Nitride Fuels
• 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
• Evaluation Methods for Corrosion Damage of Components in Cooling Systems of Nuclear Power Plants by Coupling Analysis of Corrosion and Flow Dynamics (I) : Major Targets and Development Strategies of the Evaluation Methods
• A Novel Echocardiographic Classification of Left Ventricular Diastolic Dysfunction in Patients with Preserved Systolic Function
• Relation between Mitral Annular Calcification and Severity of Carotid Arterosclerosis in Stroke Patients
• Calculation of Heat Transfer Coefficients on a Flat Plate by Pseudo Direct Numerical Simulation of Turbulence
• Investigation of Cell Resistance for Molten Salt Electrorefining of Spent Nuclear Fuel
• PRA-Based SMA : the First Tool toward a Risk-Informed Approach to the Seismic Design of the IRIS
• Experimental and Numerical Study of Liquid Metal Boiling in a System of Parallel Bundles under Natural Circulation Conditions
• Prediction of Release Rate of Burnt Sodium as Aerosol
• International Symposium on Global Environment and Nuclear Energy Systems October 24-27, 1994, Susono Kenshu Center, Shizuoka, Japan
• 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
• Analytical Study on Detailed Void Distributions Inside BWR Fuel Bundle under Turbine Trip Event Considering Time-Dependent Pin Power Distributions
• 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
• Preface
• Numerical Prediction of Secondary Flows in Complex Areas Using Concept of Local Turbulent Reynolds Number
• Analysis of fuel pin behavior under slow-ramp type transient overpower condition by using the fuel performance evaluation code 'FEMAXI-FBR'
• High-throughput screening identified disease-causing-mutants and functional variants of α-galactosidase A gene in Japanese male hemodialysis patients

もっと見る 閉じる

スポンサーリンク