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
スポンサーリンク
概要
- 論文の詳細を見る
Two extended bias factor methods, the LC and PE methods, were applied to the prediction accuracy evaluation of neutronic characteristics of a breeding light water reactor, using data of FCA-XXII-1 critical experiments, in order to investigate the features and effectiveness of these methods on the basis of an actual core design and existing experimental results. The present study confirms the following features of these methods. Both the LC and PE methods can improve the prediction accuracy the most when all the experimental results are used. The prediction accuracy improvement is achieved mainly by reducing uncertainty due to errors in cross sections. This is done by realizing a profile of sensitivity coefficients closer to that of the target core and suppressing the influence of errors in experiments and experimental analysis methods. The PE method always improves the prediction accuracy with the use of any combination of experimental results. It is always superior to the LC method in the improvement of the prediction accuracy. Concerning the effectiveness of using the extended bias factor methods with the data of FCA XXII-1 critical experiments, it is concluded that the experimental results regarding multiplication factor are more effective than the other experimental results, namely, reaction rate ratios of 238U capture to 239Pu fission (C28/F49) and void reactivity, in reducing prediction uncertainties of all the neutronic characteristics of the target core investigated: the multiplication factor, the C28/F49, and the void reactivity of the target core. This is due to the fact that the extended bias factor methods cannot fully utilize the potential that these experimental results have for the reduction of the uncertainties due to the errors in cross sections because of their strong correlations to the target core characteristics. This failure is due to large errors in the experiments and/or the experimental analysis methods.
- 社団法人 日本原子力学会の論文
- 2008-04-25
著者
-
MORI Takamasa
Nuclear Science and Engineering Directorate, Japan Atomic Energy Agency
-
OKAJIMA Shigeaki
Nuclear Science and Engineering Directorate, Japan Atomic Energy Agency
-
Mori Takamasa
Nuclear Science And Engineering Directorate Japan Atomic Energy Agency
-
Okajima Shigeaki
Nuclear Science And Engineering Directorate Japan Atomic Energy Agency
-
TAKEDA Toshikazu
Department of Urology, Keio University School of Medicine
-
KUGO Teruhiko
Japan Atomic Energy Agency
-
Kojima Kensuke
Nuclear Science And Engineering Directorate Japan Atomic Energy Agency
-
KUGO Teruhiko
Nuclear Science and Engineering Directorate, Japan Atomic Energy Agency
-
ANDOH Masaki
Nuclear Science and Engineering Directorate, Japan Atomic Energy Agency
-
FUKUSHIMA Masahiro
Nuclear Science and Engineering Directorate, Japan Atomic Energy Agency
-
NAKANO Yoshihiro
Nuclear Science and Engineering Directorate, Japan Atomic Energy Agency
-
KITADA Takanori
Department of Sustainable Energy and Environmental Engineering, Graduate School of Engineering, Osak
-
Takeda Toshikazu
Div. Of Sustainable Energy And Enviromental Engineering Graduate School Of Engineering Osaka Univ.
-
Takeda Toshikazu
Department Of Nuclear Engineering Graduate School Of Engineering Osaka University
-
Takeda Toshikazu
Department Of Sustainable Energy And Environmental Engineering Graduate School Of Engineering Osaka
-
Kugo Teruhiko
Nuclear Science And Engineering Directorate Japan Atomic Energy Agency
-
Andoh Masaki
Nuclear Science And Engineering Directorate Japan Atomic Energy Agency
-
Kitada Takanori
Department Of Sustainable Energy And Environmental Engineering Graduate School Of Engineering Osaka
-
Kitada Takanori
Department Of Nuclear Engineering Graduate School Of Engineering
-
Okajima Shigeaki
Nuclear Sci. And Engineering Directorate Japan Atomic Energy Agency
-
Kugo Teruhiko
Nuclear Sci. And Engineering Directorate Japan Atomic Energy Agency
-
Takeda Toshikazu
Division Of Sustainable Energy And Environmental Engineering Graduate School Of Engineering Osaka Un
-
Fukushima Masahiro
Nuclear Science And Engineering Directorate Japan Atomic Energy Agency
-
Nakano Yoshihiro
Nuclear Science And Engineering Directorate Japan Atomic Energy Agency
-
Takeda Toshikazu
Department of Nuclear Energy, Graduate School of Engineering, Osaka University
関連論文
- Analysis of the Main Steam Line Break Benchmark (Phase II) Using ANCK/MIDAC Code
- Simple and Efficient Parallelization Method for MOC Calculation
- Measurement and Analysis of Reactivity Worth of ^Np Sample in Cores of TCA and FCA
- JENDL-4.0 benchmarking for fission reactor applications
- JENDL-4.0 Benchmarking for Fission Reactor Applications
- Brain metastasis of a papillary renal cell carcinoma, identified as type 2
- Analysis of the SPERT-III E-Core Using ANCK Code with the Chord Weighting Method
- The Verification of 3 Dimensional Nodal Kinetics Code ANCK Using Transient Benchmark Problems
- Effect of Polynomial Expansion Order of Intranode Flux Treatment in Nodal SN Transport Calculation Code NSHEX for Large-Size Fast Power Reactor Core Analysis
- 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
- Measurement and Analysis of ^Am Fission Rate Ratio Relative to ^U Fission Rate in Thermal Neutron Systems Using Kyoto University Critical Assembly
- Measurement of^ Np Fission Rate Ratio Relative to ^U Fission Rate in Cores with Various Thermal Neutron Spectrum at the Kyoto University Critical Assembly
- 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
- Depletion Calculations for PWR Assemblies including Burnable Absorbers with Lattice Code PARAGON
- Spatially Dependent Self-Shielding Method with Temperature Distribution for the Two-Dimensional Transport Code PARAGON
- Development of Spatially Dependent Resonance Shielding Method
- Applicability of Constant Flux Approximation in Method of Characteristics with Filtering to Tiny Regions
- 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
- Development of Intelligent Code System to Support Conceptual Design of Nuclear Reactor Core
- Measurement and Analysis of Reactivity Worth of 241Am Sample in Water-Moderated Low-Enriched UO2 Fuel Lattices at TCA
- Minor Actinides Incineration by Loading Moderated Targets in Fast Reactor
- 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
- 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
- Measurement and Analysis of ^U Doppler Reactivity Effect in FCA Cores Simulating Light-Water-Moderated MOX Fuel Lattices
- Conceptual Design of Innovative Water Reactor for Flexible Fuel Cycle (FLWR) and its Recycle Characteristics
- Sensitivity Analysis for Multiplication Factor Change of LWR Cell Caused by the Differences between JENDL-3.2 and JENDL-3.3
- JENDL-4.0 Benchmarking for Fission Reactor Applications
- Measurement and Analysis of Reactivity Worth of ^Am Sample in Water-Moderated Low-Enriched UO_2 Fuel Lattices at TCA
- Effect of Polynomial Expansion Order of Intranode Flux Treatment in Nodal S_N Transport Calculation Code NSHEX for Large-Size Fast Power Reactor Core Analysis
- 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
- Impact of Perturbed Fission Source on the Effective Multiplication Factor in Monte Carlo Perturbation Calculations
- Effective Spatial Homogenization with Neutron Leakage Effect for FBR Control Rods
- Application of Neural Network to Multi-Dimensional Design Window Search in Reactor Core Design
- 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.
- Spatially Dependent Self-Shielding Method with Temperature Distribution for the Two-Dimensional Transport Code PARAGON
- 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.