Effect of Radial Void Distribution within Fuel Assembly on Assembly Neutronic Characteristics

スポンサーリンク

概要

• 論文の詳細を見る

• The effect of radial subchannel-wise void distribution in a fuel assembly on assembly neutronic characteristics has been investigated using the assembly calculation code SRAC95 and the subchannel analysis code THERMIT2. With the iterative calculation of assembly calculation and the subchannel analysis (Method 1), subchannel-wise void fraction distribution, pin-power distribution and the infinite multiplication factor of the assembly are calculated. The results are compared with the result of the assembly calculation using uniform void distribution as input (Method 2). The calculation is performed for two assembly configurations in the present study: one is a fuel assembly that does not include a water rod (Case 1) and the other is the assembly that includes a water rod (Case 2). The differences in the infinite multiplication factor and pin-power peaking factor between the two methods are small in both cases. In typical BWR fuel assemblies that are investigated in the present study, the method that does not consider the radial subchannel-wise void fraction distribution within a fuel assembly (Method 2) is accurate enough for practical applications.

• 社団法人 日本原子力学会の論文
• 2002-01-25

著者

• TAKEDA Toshikazu

  Department of Urology, Keio University School of Medicine

• Takeda T

  Osaka Univ. Osaka

• Takeda Toshikazu

  Department Of Nuclear Engineering Graduate School Of Engineering Osaka University

• AMA Tsuyoshi

  Department of Nuclear Engineering, Graduate School of Engineering, Osaka University

• HYOUDOU Hideaki

  Department of Nuclear Engineering, Graduate School of Engineering, Osaka University

• Ama Tsuyoshi

  Department Of Nuclear Energy Graduate School Of Engineering Osaka University

• Takeda T

  Division Of Sustainable Energy And Environmental Engineering Graduate School Of Engineering Osaka Un

• Takeda Toshikazu

  Nuclear Engineering Department Osaka University

• Hyoudou Hideaki

  Department Of Nuclear Energy Graduate School Of Engineering Osaka University

• Takeda Toshikazu

  Department of Nuclear Energy, Graduate School of Engineering, Osaka University

関連論文

• Brain metastasis of a papillary renal cell carcinoma, identified as type 2
• Prediction Accuracy Improvement of Neutronic Characteristics of a Breeding Light Water Reactor Core by Extended Bias Factor Methods with Use of FCA-XXII-1 Critical Experiments
• Rapid Estimation of Core-Power Ratio in Coupled-Core System by Rod Drop Method
• Analysis of First-Harmonic Eigenvalue Separation Experiments on KUCA Coupled-Core
• Improvement of Fitting Method of Multiband Parameters for Cell Calculations
• Space and Angular Dependence of Interface Currents in the Multiband-CCCP Method
• Theoretical Study on New Bias Factor Methods to Effectively Use Critical Experiments for Improvement of Prediction Accuracy of Neutronic Characteristics
• Effect of Moderator Density Distribution of Annular Flow on Fuel Assembly Neutronic Characteristics in Boiling Water Reactor Cores
• Nonlinear Behavior under Regional Neutron Flux Oscillations in BWR Cores
• Spatial-harmonic Neutron Spectrum Effect on Frequency-domain Modal Analysis of Regional Stability in BWR
• Heat Transfer Study for Thermal-Hydraulic Design of the Soild-Target of Spallation Neutron Source
• Evaluation of Prediction Error Reduction for Breeding Ratio by Bias Factor Method with Use of Experimental Result on Basic Reaction Rate Ratio
• Minor Actinides Incineration by Loading Moderated Targets in Fast Reactor
• Development an Verification of and Efficient Spatial Neutron Kinetics Method for Reactivity-Initiated Event Analyses
• Effective Convergence of Fission Source Distribution in Monte Carlo Simulation
• Effect of Radial Void Distribution within Fuel Assembly on Assembly Neutronic Characteristics
• Reaction Rate Calculation in Fast Reactor Blanket Using Multiband S_n Theory
• Effective Cross Section of ^U Samples for Analyzing Doppler Effect Measurements in Fast Critical Assembly
• Three Year Experimental Study on Leaching Behavior from Low Level Radioactive Plastic-Based Waste
• Evaluation of Eigenvalue Separation by the Monte Carlo Method
• Neutron Anisotropic Scattering Effect in Heterogeneous Cell Calculations of Light Water Reactors
• Estimation of Error Propagation in Monte-Carlo Burnup Calculations
• Transport Calculations of MOX and UO_2 Pin Cells by the Method of Characteristics
• Parametric Study on Fast Reactors with Low Sodium Void Reactivity by the Use of Zirconium Hydride Layer in Internal Blanket
• Sensitivity Analysis for Multiplication Factor Change of LWR Cell Caused by the Differences between JENDL-3.2 and JENDL-3.3
• Correlation between Cumulative Leaching Ratio and pH of Leachate from Cementation and Plastic Solidified Radwaste Form.
• New Control Rod Homogenization Method for Fast Reactors
• An Improvement of the Transverse Leakage Treatment for the Nodal S_N Transport Calculation Method in Hexagonal-Z Geometry
• Effective Spatial Homogenization with Neutron Leakage Effect for FBR Control Rods
• Transumbilical approach for laparo-endoscopic single-site adrenalectomy : Initial experience and short-term outcome
• Extension of fission product model for use in lattice calculation of thorium fueled BWR.
• Neutron Anisotropic Scattering Effect in Heterogeneous Cell Calculations of Light Water Reactors
• Unified diffusion coefficient for analysis of sodium-void worth in fast critical assembly with control-rod channels.
• Effect of neutron leakage in cell calculations of fast critical assembly.
• Determination of effective diffusion parameters in thermal reactor assemblies.
• Monte-Carlo/Collision Probability Hybrid Method for LWR Fuel Assembly Burnup Calculations.
• Application of improved coarse mesh method to BWR core calculations.
• Application of depletion perturbation theory to fuel loading optimization.
• Approximate Calculation Method for Second Order Sensitivity Coefficient.
• Three-dimensional transport calculation method for eigenvalue problems using diffusion synthetic acceleration.
• Three-dimensional transport correction in fast reactor core analysis.
• Development of three-dimensional transport and diffusion codes based on nodal method.
• Fast Reactor of Negativ Na Void Reactiv and Its Transient Behavior.
• A New Nodal SN Transport Method for Three-Dimensional Hexagonal Geometry.
• Determination of cell averaged diffusion constants based on transport/diffusion perturbation theory.

もっと見る 閉じる

スポンサーリンク