研究者詳細 - 山本　章夫

論文 - 山本　章夫

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Inverse estimation methods of unknown radioactive source for fuel debris search

Sugaya S., Endo T., Yamamoto A.

Annals of Nuclear Energy 124 巻頁： 49 - 57 2019年2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

To identify the distribution of fuel debris remaining in the reactor vessel and/or the containment vessel of Fukushima Daiichi NPS, we focused on the inverse estimation of radioactive source distribution using the measured radiation counts. The Maximum Likelihood-Expectation Maximization (ML-EM) and the Moore-Penrose Matrix Inverse (MPMI) methods are examined. The ML-EM method has been used for the image reconstruction of computed tomography, and the MPMI method is one of the solution methods for simultaneous linear equations with the underdetermined condition. A simple calculation model simulating a containment vessel was constructed including detectors and radiation sources. In an actual situation, a sufficient number of radiation measurement positions would not be available owing to the complexity of structures inside the containment vessel. Thus, the number of radiation measurement points (number of constraints) is smaller than that of radiation source positions. It means that an underdetermined inverse problem should be solved. The detection probability of radiation (neutron or photon) is calculated by the adjoint transport calculation since the detection probability is used as the coupling coefficient between radiation counts at a detector and a radioactive source. The result of estimation using the ML-EM or the MPMI method indicates that the accuracy of estimation depends on the distance between a radiation source and a detector, and measurement positions of radiation count. The ML-EM and the MPMI methods show different prediction accuracy depending on the prediction condition. It is found that reasonable prediction accuracy would be obtained when the detectors are placed at the vicinity of radiation sources of interest.

DOI： 10.1016/j.anucene.2018.09.022

Web of Science

Scopus

Comparison of theoretical formulae and bootstrap method for statistical error estimation of Feynman-α method

Endo T., Yamamoto A.

Annals of Nuclear Energy 124 巻頁： 606 - 615 2019年2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

This paper discusses the statistical error of the variance-to-mean ratio, or the Y value in the Feynman-α method, from a single measurement of reactor noise. As a theoretical approach, two practical theoretical formulae are derived to estimate the statistical error of Y: one is based on the propagation of uncertainty with unbiased estimators for the third- and fourth-order central moments; the other is a simplified formula that reuses the Y value under the fundamental mode approximation, where the subcriticality is approximately less than 10,000 pcm. As a numerical approach, the bootstrap method is improved to efficiently estimate the correlations of Y between different counting gate widths, or covariance matrix ΣY, due to the bunching method. Through an actual reactor noise experiment at the Kyoto University Criticality Assembly, the statistical errors of Y using the theoretical formulae and the bootstrap method are validated by comparing the reference statistical errors that are estimated from the multiple experiments of reactor noise. Furthermore, the impact of ΣY on the statistical error of the prompt neutron decay constant α is numerically investigated. Consequently, in the case of this experimental analysis, it was confirmed that the bootstrap method with the correlations of Y seems to be slightly better from the viewpoint of the coverage probability of the estimated confidence intervals of α although the fitting error method without the correlation of Y could be useful for the order estimation of the statistical error of α.

DOI： 10.1016/j.anucene.2018.10.032

Web of Science

Scopus

A New Interpretation of Discontinuity Factor

Yamamoto A., Endo T.

Nuclear Science and Engineering 193 巻 ( 9 ) 頁： 991 - 997 2019年

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

A new interpretation of the discontinuity factor (DF) for scalar flux, partial current, and angular flux is discussed. Conventionally, the DF is considered as the discontinuous condition of scalar flux, partial current, or angular flux at an interface. In the new interpretation, the DF is considered as the refractive index of materials for partial current or angular flux that conserves odd-parity or odd-moment angular flux at an interface of different materials. It is related to the transmission and reflection of partial current or angular flux at an interface where different materials are adjacent. Using the present interpretation, a fundamental issue of neutron balance (i.e., artificial loss or production of neutrons at an interface due to discontinuous condition), which would appear in the conventional interpretation of DF, can be resolved.

DOI： 10.1080/00295639.2019.1579514

Web of Science

Scopus

DEVELOPMENT OF ASSEMBLY BOWING MODEL FOR PIN-BY-PIN CORE CALCULATIONS

Yamamoto Kento, Ohoka Yasunori, Nagano Hiroaki, Yamamoto Akio, Endo Tomohiro

The Proceedings of the International Conference on Nuclear Engineering (ICONE) 2019.27 巻 ( 0 ) 頁： 1022 2019年

記述言語：英語掲載種別：研究論文（学術雑誌）出版者・発行元：The Japan Society of Mechanical Engineers

Calculation capability of the pin-power distribution considering the variation of the assembly gap size due to assembly bowing was implemented in the pin-by-pin core calculation code SCOPE2. The previous studies show that the perturbation of geometry can be treated by the perturbation of macroscopic cross-section or atomic number density of the region instead of explicit consideration of geometry deformation. This methodology was applied to the assembly gap region: the variation of the gap size was treated by the correction on the macroscopic cross-section of the gap water. The correction model for the cross-section of gap water was implemented in SCOPE2 since it treats pin-by-pin cross-sections in a transport calculation. The correction was made according to the variation in the gap size. This implemented model has an advantage that the modification of the cross-section tables used in the core calculation is not necessary to consider the variation of gap size. For single assembly and multi-assemblies geometries, the assembly bowing model implemented in SCOPE2 was verified by comparing with the reference results using the assembly calculation code AEGIS, where the gap size perturbation was explicitly considered by varying the geometry of the gap region. It was confirmed that the variation of pin-power distribution due to the assembly bowing can be appropriately treated by SCOPE2 with the assembly bowing model.

DOI： 10.1299/jsmeicone.2019.27.1022

CiNii Research

基礎から分かる未臨界

遠藤知弘, 辻本和文, 山本章夫

日本原子力学会誌ATOMOΣ 61 巻 ( 10 ) 頁： 734 - 738 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：一般社団法人日本原子力学会

<p>　臨界管理や原子炉の運転等において，体系の中性子実効増倍率は最も基本的な概念の一つである。実効増倍率が1のときを臨界，1未満のときを未臨界と呼ぶことは良く知られているが，実は両者は大きく異なる状態である。本連載講座では，核燃料サイクル施設の臨界安全，原子炉施設の臨界管理や福島第一原子力発電所のデブリ取り出し等で重要となる未臨界状態について，臨界状態の原子炉との違い，炉物理的な特徴，未臨界状態であることを評価・測定する手法などを解説する。第1回では，臨界・未臨界の炉物理的な違いについて紹介する。</p>

DOI： 10.3327/jaesjb.61.10_734

CiNii Research

炉物理の使命

千葉豪, 山本章夫

日本原子力学会誌ATOMOΣ 61 巻 ( 4 ) 頁： 254 - 256 2019年

Uncertainty quantification/reduction of BWR core characteristics considering cross section and thermal-hydraulics uncertainties

Ito M., Yamamoto A., Endo T., Ama T.

Transactions of the American Nuclear Society 120 巻頁： 851 - 854 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

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

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

Resonance calculation using energy spectral expansion based on reduced order model: Application to heterogeneous geometry

Yamamoto A., Kondo R., Endo T., Takeda S., Koike H., Yamaji K., Sato D.

Transactions of the American Nuclear Society 121 巻頁： 1316 - 1320 2019年

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

A resonance calculation method using energy expansion bases based on a reduced order model

Yamamoto A., Endo T., Takeda S., Koike H., Yamaji K., Ieyama K., Sato D.

International Conference on Mathematics and Computational Methods Applied to Nuclear Science and Engineering, M and C 2019 頁： 1081 - 1092 2019年

A Resonance calculation using energy Spectral Expansion bases (the RSE method) is newly proposed. In the present method, an ultra-fine group spectrum appeared in a resonance calculation is expanded by expansion bases on energy. The transport equation for the expansion coefficients is derived, which is the simultaneous first order differential equation for the expansion coefficients. The present method is applied to a one-dimensional slab geometry composed of 238U. Comparisons of the results obtained by the present and the ultra-fine group (reference) methods show the fundamental validity of this method.

Scopus

Application of various superhomogenization (SPH) methods for the method of characteristics

Sawada K., Endo T., Yamamoto A.

International Conference on Mathematics and Computational Methods Applied to Nuclear Science and Engineering, M and C 2019 頁： 1534 - 1543 2019年

To reduce the error due to the difference between the fine and the coarse calculation conditions in a transport calculation using method of characteristics (MOC), the superhomogenization (SPH) method has been widely used. However, the conventional SPH method has a problem with numerical stability in the iterative calculations for the estimation of SPH factors, especially when local strong absorbers exist in a calculation geometry. To address this problem, improvements of the SPH methods were tested to reduce the angular and spatial discretization error of the MOC. The performances of the various SPH methods were confirmed in pressurized water reactor (PWR) 2 × 2 fuel assembly geometries and in a PWR core geometry. The results of verification indicate that in a core geometry the improved SPH methods provide better accuracy than that of the conventional SPH.

Scopus

Calculation method of estimated criticality lower-limit multiplication factor using the bootstrap method

Hayashi T., Endo T., Yamamoto A.

International Conference on Mathematics and Computational Methods Applied to Nuclear Science and Engineering, M and C 2019 頁： 1874 - 1885 2019年

The present paper describes a calculation method of the estimated criticality lower-limit multiplication factor (ECLLMF) using the bootstrap method, which is an uncertainty evaluation method. In the conventional method, the normal distribution is assumed for a probability distribution with respect to calculation biases of effective multiplication factors (keff) to evaluate ECLLMF. To propose a methodology without the assumption of the normal distribution, an estimation method of ECLLMF is newly developed using the bootstrap method, where the assumption of the normal distribution is not necessary. From the verification results, it is confirmed that the estimation result using the non-parametric bootstrap method is more reasonable than that using the conventional method, when a probability distribution of keff does not obey the normal distribution.

Scopus

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

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

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

Flux Region Assignment Method Using Ray Trace Information for the Method of Characteristics to Improve Cache Efficiency

Yamamoto A., Giho A., Endo T.

Nuclear Science and Engineering 192 巻 ( 3 ) 頁： 240 - 253 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：Nuclear Science and Engineering

A flux region assignment algorithm to increase cache efficiency for the method of characteristics (MOC) is proposed. In order to minimize the stride of memory access, flux region identifications are assigned based on the ray trace sequence during the MOC calculation. The present method is implemented in the three-dimensional transport code GENESIS and its performance is confirmed through verification calculations ranging from single pressurized water reactor (PWR) fuel assembly to PWR full-core benchmark problems. Quantitative comparison of cache efficiency is carried out and the present method shows improved cache efficiency, which results in a reduction in computation time. The present method can reduce computational time by improving cache efficiency while suppressing memory requirement.

DOI： 10.1080/00295639.2018.1501978

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

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

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