Takeda Toshikazu | Department of Nuclear Energy, Graduate School of Engineering, Osaka University

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著作論文

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
Minor Actinides Incineration by Loading Moderated Targets in Fast Reactor
Effect of Radial Void Distribution within Fuel Assembly on Assembly Neutronic Characteristics
Effective Cross Section of ^U Samples for Analyzing Doppler Effect Measurements in Fast Critical Assembly
Evaluation of Eigenvalue Separation by the Monte Carlo Method
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
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.

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