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論文 - 山本　章夫

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Data assimilation using subcritical measurement of prompt neutron decay constant

Endo T., Yamamoto A.

International Conference on Mathematics and Computational Methods Applied to Nuclear Science and Engineering, M and C 2019 頁： 1864 - 1873 2019年

記述言語：日本語掲載種別：研究論文（学術雑誌）出版者・発行元：International Conference on Mathematics and Computational Methods Applied to Nuclear Science and Engineering, M and C 2019

The prompt neutron decay constant α can be directly measured by the reactor noise analysis method (e.g., the Feynman-α method) in a steady-state subcritical system. In this study, the applicability of the data assimilation techniques (i.e., the bias factor and cross section adjustment methods) using a subcritical measurement of α conducted at the Kyoto University Critical Assembly (KUCA) was investigated to reduce the nuclear data-induced uncertainty of keff.The sensitivity coefficients of keff and α with respect to the nuclear data were efficiently estimated using a deterministic SN transport code with the first-order perturbation theory. As a result, a priori relative uncertainty of keff due to the 56 group SCALE covariance data can be reduced. The experimental value of α contributes to improving nuclear data of fission spectrum X and total fission neutron number v via strong correlations between X and prompt Xp and between v and prompt vp.

Scopus

Development of efficient data sampling method to construct surrogate model of severe accident analysis code for SBO aiming probabilistic safety margin analysis

Matsushita M., Endo T., Yamamoto A.

Transactions of the American Nuclear Society 121 巻頁： 979 - 982 2019年

Implementation of random sampling for ace-format cross sections using frendy and application to uncertainty reduction

Kondo R., Endo T., Yamamoto A., Tada K.

International Conference on Mathematics and Computational Methods Applied to Nuclear Science and Engineering, M and C 2019 頁： 1493 - 1502 2019年

For uncertainty quantification and reduction using the random sampling technique through a continuous energy Monte Carlo method, the perturbation capability of nuclear data libraries for MCNP is developed using a nuclear data processing system FRENDY, which is being developed by JAEA. The implemented capability is applied to uncertainty quantification using the random sampling method and to uncertainty reduction of kinetic parameter based on the bias factor method. Validity of the implemented capability is confirmed through comparison with the results obtained by the conventional sandwich formula using SCALE and MCNP.

Scopus

Reduction of macroscopic and microscopic cross section table size for heterogeneous core calculation using dimensionality reduction

Yamamoto M., Endo T., Yamamoto A.

Transactions of the American Nuclear Society 121 巻頁： 1309 - 1312 2019年

Underestimation of statistical uncertainty of local tallies in Monte Carlo eigenvalue calculation for simple and LWR lattice geometries

Hayashi, K; Endo, T; Yamamoto, A

JOURNAL OF NUCLEAR SCIENCE AND TECHNOLOGY 55 巻 ( 12 ) 頁： 1434 - 1458 2018年12月

記述言語：英語掲載種別：研究論文（学術雑誌）出版者・発行元：Journal of Nuclear Science and Technology

A prediction method of the true variance of local tally in a Monte Carlo (MC) critical calculation is developed. In the MC calculation, the effective multiplication factor ((Formula presented.)) and the fission rate distribution are estimated by simulating fission chain reactions. The statistical uncertainty of calculation result is commonly estimated by the standard error based on the central limit theorem. However, the evaluated statistical uncertainty would be underestimated when the inter-cycle correlation is not appropriately taken into account. In this study, a theoretical formula of the underestimation ratio (UR) of the statistical uncertainty for a local tally is derived using the eigenfunction expansion and the Autoregressive model. Note that the UR is defined by the ratio of uncertainty estimated by a MC calculation to the true statistical uncertainty. The proposed method is applied to one-dimensional slab and multi-assembly geometries with reflective boundary conditions. In the one-dimensional slab geometry, the prediction results of UR show reasonable agreement with the reference. In the multi-assembly geometry, on the other hand, the prediction results of UR are agreement with the reference regarding the relative spatial shape but there are considerable differences in the absolute values of UR.

DOI： 10.1080/00223131.2018.1513875

Web of Science

Scopus

Subchannel void distribution correction model for the two-stage core analysis method in boiling water reactors

Mitsuyasu, T; Aoyama, M; Yamamoto, A

ANNALS OF NUCLEAR ENERGY 122 巻頁： 146 - 154 2018年12月

記述言語：英語掲載種別：研究論文（学術雑誌）出版者・発行元：Annals of Nuclear Energy

The two-stage core analysis method is widely used for BWR core analysis. The purpose of this study is to develop a subchannel void distribution correction model for the two-stage core analysis method using an assembly-based thermal-hydraulics calculation in the core analysis stage. The model assumes two kinds of subchannel void distribution gradients along with the two diagonal lines in the horizontal cross section of a BWR fuel assembly. The model appends and tabulates the difference of the subchannel perturbation condition from the base condition in the lattice physics, and evaluates the tilts within the 2D lattice physics scheme, and couples those results with 3D subchannel analysis which evaluates the thermal-hydraulics characteristics within the coolant flow area divided as some subchannel regions. The developed model is evaluated using a heterogeneous and a small core problem. The model gives a better power distribution compared with that of the authors’ previous model. As a result, the model can incorporate the subchannel effect into the current two-stage core calculation method.

DOI： 10.1016/j.anucene.2018.08.041

Web of Science

Scopus

Sensitivity analysis of prompt neutron decay constant using perturbation theory

Endo, T; Yamamoto, A

JOURNAL OF NUCLEAR SCIENCE AND TECHNOLOGY 55 巻 ( 11 ) 頁： 1245 - 1254 2018年11月

記述言語：英語掲載種別：研究論文（学術雑誌）出版者・発行元：Journal of Nuclear Science and Technology

Experimental results of prompt neutron decay constant α is useful information to validate numerical results of ω-eigenvalue for spatial and energetic fundamental mode. In order to accomplish the data assimilation technique using α, it is desirable to establish an efficient numerical calculation method for sensitivity coefficient analysis of α For this purpose, the numerical calculation method using the first-order perturbation theory is investigated. A specific theoretical formula is derived to evaluate the sensitivity coefficient of α to nuclear data. The derived rigorous formula utilizes forward and adjoint eigenfunctions which consist of neutron flux and delayed neutron precursor densities. Using the prompt approximation, the derived formula can be simplified without the term involving the delayed neutron precursor densities. By calculating α using the multi-energy-group neutron transport code for an ICSBEP benchmark problem, the derived formula for sensitivity analysis using the perturbation theory is verified by comparing with the reference results using the direct method. Consequently, the efficient numerical procedures for uncertainty quantification of α can be established by the aid of the sensitivity coefficients based on the perturbation theory.

DOI： 10.1080/00223131.2018.1491902

Web of Science

Scopus

Reduction of Cross Section Table Size for Core Analysis Using Dimensionality Reduction Technique 査読有り

M. Yamamoto, T. Endo, A. Yamamoto

Trans. Am. Nucl. Soc. 119 巻頁： 1226-1228 2018年11月

A Simple Treatment of Bowed Assembly Gap Through Correction of Cross Section 査読有り

A. Yamamoto, T. Endo, K. Yamamoto, Y. Ohoka, H. Nagano

Trans. Am. Nucl. Soc. 119 巻頁： 1199-1202 2018年11月

Transport Consistent Diffusion Coefficient for CMFD Acceleration 査読有り

A. Yamamoto, T. Endo

Trans. Am. Nucl. Soc. 119 巻頁： 1179-1181 2018年11月

Estimation of Subcriticality in Dollar Units Based on Integral Method for Arbitrary State-Change in Subcritical System 査読有り

A. Nonaka, T. Endo, A. Yamamoto

Trans. Am. Nucl. Soc. 119 巻頁： 1112-1115 2018年11月

Surrogate Model of Severe Accident Analysis Code for SBO Aiming Probabilistic Safety Margin Analysis 査読有り

M. Matsushita, T. Endo, A. Yamamoto

Trans. Am. Nucl. Soc. 119 巻頁： 900-903 2018年11月

Estimation of Subcriticality using Particle Filter Method 査読有り

T. Ikeda, T. Kimura, T. Endo, A. Yamamoto

Trans. Am. Nucl. Soc. 118 巻頁： 851-854 2018年6月

Quantification of Modeling Approximation Error of Pin-Cell Calculation Using Kriging and Principal Component Analysis 査読有り

T.Hanai, A. Yamamoto, T. Endo, K. Yamamoto, Y. Ohoka, H. Nagano

Trans. Am. Nucl. Soc. 118 巻頁： 875－878 2018年6月

Inverse Estimation Methods of Unknown Radioactive Source for Fuel Debris Search 査読有り

S. Sugaya, T. Endo, A. Yamamoto

Proc. PHYSOR201 頁： [USB-DRIVE] 2018年4月

Estimation of Subcriticality in Dollar Units using Integral Method for Subcritical System 査読有り

A. Nonaka, T. Endo, A. Yamamoto

Proc. PHYSOR2018 頁： [USB-DRIVE] 2018年4月

Cache Efficient Flux Region Assignment for the Method of Characteristics 査読有り

A. Yamamoto, A. Giho, T. Endo

Proc. PHYSOR2018 頁： [USB-DRIVE] 2018年4月

Estimation of Region-Wise Even-Parity Discontinuity Factor for MOC Through Iterative Procedure 査読有り

A. Yamamoto, A. Giho, T. Endo

Proc. PHYSOR2018 頁： [USB-DRIVE] 2018年4月

Development of Reduced Order Model of Severe Accident Analysis Code for Probabilistic Safety Margin Analysis

M. Matsushita, T. Endo, A. Yamamoto, T. Kitao

Proc. PHYSOR2018 頁： [USB-DRIVE] 2018年4月

Sensitivity Coefficient Analysis of Omega-Eigenvalue based on First-Order Perturbation Theory 査読有り

頁： [USB-DRIVE] 2018年4月

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