研究者詳細 - 山本　章夫

論文 - 山本　章夫

分割表示 446 件中 21 - 40 件目／全件表示 >>

Applicability of Dynamic Mode Decomposition to Estimate Fundamental Mode Component of Prompt Neutron Decay Constant from Experimental Data

Nishioka F., Endo T., Yamamoto A., Yamanaka M., Pyeon C.H.

Nuclear Science and Engineering 196 巻 ( 2 ) 頁： 133 - 143 2022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

To robustly estimate the fundamental mode component of prompt neutron decay constant α in a subcritical system, dynamic mode decomposition (DMD) is applied to time-series data obtained by the pulsed-neutron source (PNS) and Rossi-α methods. For the statistical uncertainty quantification of α by DMD, randomly sampled virtual data are used for the DMD procedure. The applicability of DMD is demonstrated by analyzing the experimental results by the PNS and Rossi-α methods, which are performed at the Kyoto University Critical Assembly (KUCA). When applying the DMD to the PNS and Rossi-α experimental data, a constant signal was added to the experimental data to remove the background constant component. The application results indicate that DMD enables one to robustly estimate the fundamental mode component of α in the PNS and Rossi-α methods.

DOI： 10.1080/00295639.2021.1968225

Application of dynamic mode decomposition to Rossi-α method in a critical state using file-by-file moving block bootstrap method

Endo T., Nishioka F., Yamamoto A., Watanabe K., Pyeon C.H.

Journal of Nuclear Science and Technology 59 巻 ( 9 ) 頁： 1117 - 1126 2022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：Journal of Nuclear Science and Technology

Prompt neutron decay constant (Formula presented.) in a critical state is useful information to validate the numerically predicted ratio of the point kinetics parameters (Formula presented.), where (Formula presented.) and (Formula presented.) are the effective delayed neutron fraction and prompt neutron lifetime, respectively. To directly measure (Formula presented.) in a target critical system, this study proposes the application of the dynamic mode decomposition (DMD) to the reactor noise analysis based on the Rossi- (Formula presented.) method. The DMD-based Rossi- (Formula presented.) method enables us to robustly estimate the fundamental mode component of (Formula presented.) from the Rossi- (Formula presented.) histograms measured using multiple neutron detectors. Furthermore, the file-by-file moving block bootstrap method is newly proposed for the statistical uncertainty quantification of (Formula presented.) to prevent huge memory usage when the neutron count rate is high and/or the total measurement time is long. A critical experiment has been conducted at Kyoto University Critical Assembly to demonstrate the proposed method. As a result, the proposed method can uniquely determine the (Formula presented.) value of which the statistical uncertainty is smallest. By utilizing this experimental result of (Formula presented.), numerical results of (Formula presented.) ratio using the continuous energy Monte Carlo code MCNP6.2 with recent nuclear data libraries, which are processed by the nuclear data processing code FRENDY, are validated.

DOI： 10.1080/00223131.2022.2030260

外的事象に対する原子力発電所の安全対策とリスクマネジメント

村上健太, 山本章夫

日本原子力学会誌ATOMOΣ 64 巻 ( 5 ) 頁： 272 - 274 2022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>　フォローアップセミナーでは，企画セッションで抽出された2つの論点について，講演者に2名のパネリストを加えて議論した。リスク評価を含む知見の不確かさを踏まえながら迅速な意思決定を行うという論点では，ハザード評価の確からしさに関する認識の相違が更に浮き彫りになると共に，得られた知見に対して「規制基準に照らして判断する」という議論と，不確かさを有することを前提に対策を決めることの違いが指摘された。また外的事象に対するプラントのパフォーマンスを評価するという論点では，地震PRAを十分に活用して，設計基準レベルを超えたときのプラントの挙動を把握することの重要性が強調された。</p>

DOI： 10.3327/jaesjb.64.5_272

Implementation of Resonance Upscattering Treatment in FRENDY Nuclear Data Processing System

Yamamoto A., Endo T., Chiba G., Tada K.

Nuclear Science and Engineering 196 巻 ( 11 ) 頁： 1267 - 1279 2022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

The resonance upscattering effect (the thermal agitation effect) is implemented in the generation capability of multigroup neutron cross sections of the FRENDY nuclear data processing system. The resonance upscattering effect is considered by (1) the variation of self-shielding factors (effective cross sections) due to the change in the ultra-fine group spectrum and (2) the variation of group-to-group elastic scattering cross sections. Since the upscattering effect is considered in the ultra-fine group spectrum calculation, an iteration calculation is necessary to consider the upscattering. The impacts of the iteration strategy (Jacobi or Gauss-Seidel), as well as the number of iterations, are discussed. In the verification calculations, impacts on the ultra-fine group spectrum, effective cross sections, and neutronics characteristics (the Doppler effect) are confirmed. The effect of energy group structure and the impact of resonance upscattering treatments on the Doppler effect through the variation of effective cross sections and the elastic scattering matrix are investigated. The results indicate that FRENDY can provide appropriate multigroup cross sections considering the resonance upscattering effect.

DOI： 10.1080/00295639.2022.2087833

Sensitivity Coefficient Evaluation of an Accelerator-Driven System Using ROM-Lasso Method

Katano R., Yamamoto A., Endo T.

Nuclear Science and Engineering 196 巻 ( 10 ) 頁： 1194 - 1208 2022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

We propose the use of reduced-order modeling to improve the sensitivity coefficient evaluation method based on Lasso-type penalized linear regression. In this method, cross sections of interest are uniformly randomly sampled, and corresponding perturbed core analyses are performed. The sensitivity coefficients of the higher-dimensional model are expanded by the active subspace (AS) attained by the lower-dimensional model, and the expansion coefficients are estimated by the Lasso regression. In addition, AS bases can be flexibly chosen according to neutronics parameters of interest. We conducted a verification calculation for an accelerator-driven system and clarified that the proposed method successfully reduces the calculation cost by a couple of orders of magnitude compared with the direct method. The proposed method can be used to practically evaluate the sensitivity coefficients of various parameters.

DOI： 10.1080/00295639.2022.2067447

倫理規程制定20年を迎えて　第4回

山本章夫

日本原子力学会誌ATOMOΣ 64 巻 ( 3 ) 頁： 176 - 176 2022年

組織横断的な原子力教育基盤の構築に向けて

黒﨑健, 小崎完, 中島宏, 小原徹, 若林源一郎, 卞哲浩, 松山成男, 阿部博志, 宇埜正美, 山本章夫

日本原子力学会誌ATOMOΣ 64 巻 ( 9 ) 頁： 520 - 524 2022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>　文部科学省が支援する国際原子力人材育成イニシアティブ事業において，未来社会に向けた先進的原子力教育コンソーシアム(Advanced Nuclear Education Consortium for the Future Society：ANEC)が2020年度に設立された。本稿では，ANEC設立の背景や経緯，運営体制，人材育成の取り組み状況，将来展望等を紹介する。</p>

DOI： 10.3327/jaesjb.64.9_520

Proposal and Application of ROM-Lasso Method for Sensitivity Coefficient Evaluation

Katano R., Yamamoto A., Endo T.

Proceedings of the International Conference on Physics of Reactors, PHYSOR 2022 頁： 2032 - 2041 2022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：Proceedings of the International Conference on Physics of Reactors, PHYSOR 2022

We propose a novel method for evaluating sensitivity coefficients of neutronics parameters to cross sections, so-called the ROM-Lasso. In this method, cross sections of interest are randomly sampled, and corresponding perturbed core analyses are performed. Then, the sensitivity coefficient vector of the higher-level model is expanded via the active subspace bases obtained with the lower-level model whose dimensional complexity is smaller than that of the higher-level model, and the expansion coefficients are estimated by the lasso regression. A unique feature of the ROM-Lasso method allows the use of different bases optimized for each neutronics parameter. We conducted a verification calculation for an accelerator-driven system and demonstrated that the ROM-Lasso method can reproduce the sensitivity coefficients with a much smaller number of forward calculations than the direct method. The proposed method can be used to practically evaluate the sensitivity coefficients.

DOI： 10.13182/PHYSOR22-37557

Neutron Diffusion Calculation in Heterogeneous Geometry Based on Local/Global Iteration Using Proper Orthogonal Decomposition

Ito M., Endo T., Yamamoto A., Masaoka Y., Kodama Y., Nagano H.

Proceedings of the International Conference on Physics of Reactors, PHYSOR 2022 頁： 503 - 512 2022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：Proceedings of the International Conference on Physics of Reactors, PHYSOR 2022

This study newly proposes a heterogeneous core calculation method based on local/global iteration using proper orthogonal decomposition (POD). By using the singular value decomposition (SVD) and the low-rank approximation, appropriate POD bases for expanding the neutron flux can be obtained from snapshot data of the neutron flux obtained by fine mesh calculations. By projection using the POD bases, the dimension of the target equation (e.g., discretized neutron diffusion equation) can be dramatically reduced. In the proposed method, POD is effectively applied to each single assembly calculation (local calculation). Furthermore, using the local/global iteration, the effective neutron multiplication factor and the neutron flux distribution in the whole core geometry can be obtained by combining the numerical results of the local calculation for each fuel assembly and the global calculation for the whole core. As a feasibility study, the proposed method is applied to a one-dimensional heterogeneous core analysis, and the accuracy is investigated by changing the total number of POD bases.

DOI： 10.13182/PHYSOR22-37693

Investigation of the Impact of Difference Between FRENDY and NJOY2016 on Neutronics Calculations

Ono M., Tojo M., Tada K., Yamamoto A.

Proceedings of the International Conference on Physics of Reactors, PHYSOR 2022 頁： 88 - 96 2022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：Proceedings of the International Conference on Physics of Reactors, PHYSOR 2022

A nuclear data library is used as a starting input for all subsequent neutronics calculations. NJOY has been used worldwide as a nuclear data processing code to create cross section libraries for a long time. For verification of NJOY method and providing an alternative nuclear data processing tool, JAEA has been developed the new nuclear library processing code FRENDY. In this paper, nuclear calculations were performed using the ACE files and the multigroup libraries created by both FRENDY and NJOY, and the impacts on the neutronics characteristics due to nuclear data processing were investigated using those libraries. MCNP was used to compare the ACE files by calculating many benchmark problems including ICSBEP and it was confirmed that the k-eff values generally agree with each other within the range of statistical errors. The multigroup cross sections are verified by the BWR design codes LANCR/AETNA through calculation of a commercial BWR5 equilibrium core loaded with 9x9 fuels. It was confirmed that fuel assembly and core characteristics are consistent with each other. From the above investigations, it was confirmed that FRENDY can provide comparable continuous/multi-group neutron cross sections with NJOY.

DOI： 10.13182/PHYSOR22-37287

Development of Nuclear Data Processing Code FRENDY Version 2

Tada K., Yamamoto A., Endo T., Chiba G., Ono M., Tojo M.

Proceedings of the International Conference on Physics of Reactors, PHYSOR 2022 頁： 107 - 116 2022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：Proceedings of the International Conference on Physics of Reactors, PHYSOR 2022

Nuclear data processing is an important interface between an evaluated nuclear data library and neutronics calculation codes. JAEA has been developed the new nuclear data processing code FRENDY from 2013. FRENDY version 1 generates the ACE files used for the continuous-energy Monte Carlo codes including PHITS, Solomon, Serpent, and MCNP and it was released as the open-source software under the 2-clause BSD license in 2019. After we released FRENDY version 1, many functions, e.g., the multi-group neutron cross-section library generation, the statistical uncertainty quantification of the probability tables for unresolved resonance cross-section, the perturbation of the ACE file, and the modification of the ENDF-6 formatted nuclear data file, are developed. We released FRENDY version 2 including these functions. The present paper explains the overview of FRENDY and features of new functions implemented in FRENDY version 2.

DOI： 10.13182/PHYSOR22-37299

AREA RATIO METHOD USING DYNAMIC MODE DECOMPOSITION

Endo T., Nishioka F., Yamamoto A., Yamanaka M., Pyeon C.H.

Proceedings of the International Conference on Physics of Reactors, PHYSOR 2022 頁： 3366 - 3375 2022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：Proceedings of the International Conference on Physics of Reactors, PHYSOR 2022

To reduce the higher mode effects on the subcriticality estimation using the pulsed neutron source without the aid of a high-fidelity neutron transport calculation, we propose a new area ratio method based on a data-driven approach using the dynamic mode decomposition (DMD) with the neutron count data measured by multiple neutron detectors. Thanks to eigenvalues and eigenvectors based on DMD, the fundamental mode of the prompt neutrons and the background component due to delayed neutrons are extracted from the measured neutron count data matrix. The effectiveness of the proposed method is demonstrated via experimental analysis for a pulsed neutron source measurement performed at Kyoto University Critical Assembly.

DOI： 10.13182/PHYSOR22-37184

A New Approach for Resonance Treatment of Doubly Heterogeneous Fuel using the RSE method

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

Proceedings of the International Conference on Physics of Reactors, PHYSOR 2022 頁： 2042 - 2051 2022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：Proceedings of the International Conference on Physics of Reactors, PHYSOR 2022

A new resonance calculation method for the doubly-heterogeneous (DH) fuels such as high-temperature gas-cooled reactor fuel is proposed based on the Resonance calculation based on Spectral Expansion (RSE) method. The concept of pointwise disadvantage factor for fuel grain is taken into account to treat the DH fuels. The verification calculation is carried out for simplified single fuel cell and fuel compact consisting of five fuel cells and graphite moderator. The calculation results indicate that the present method can appropriately handle the space-dependent self-shielding effect for DH fuels.

DOI： 10.13182/PHYSOR22-37251

New method for visualizing the dose rate distribution around the Fukushima Daiichi Nuclear Power Plant using artificial neural networks

Sasaki M., Sanada Y., Katengeza E.W., Yamamoto A.

Scientific Reports 11 巻 ( 1 ) 2021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：Scientific Reports

This study proposes a new method of visualizing the ambient dose rate distribution using artificial neural networks (ANNs) from airborne radiation monitoring results. The method was applied to the results of the airborne radiation monitoring which was conducted around the Fukushima Daiichi Nuclear Power Plant by an unmanned aerial vehicle. Much of the survey data obtained in the past were used as the training data for building a network. The number of training cases was related to the error between the ground and converted values by the ANN. The quantitative evaluation index (the root-mean-square error) between the ANN-converted value and the ground-based survey result converged at 200 training cases. This number of training case was considered a rough criterion of the required number of training cases. The reliability of the ANN method was evaluated by comparison with the ground-based survey data. The dose rate map created by the ANNs method reproduced ground-based survey results better than traditional methods.

DOI： 10.1038/s41598-021-81546-4

Evolutionary simulated annealing for fuel loading optimization of VVER-1000 reactor 国際共著

Tran V.P., Phan G.T.T., Hoang V.K., Ha P.N.V., Yamamoto A., Tran H.N.

Annals of Nuclear Energy 151 巻 2021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

An evolutionary simulated annealing (ESA) method has been developed for the problem of fuel loading optimization of VVER-1000 reactor. The ESA method improves original simulated annealing by using crossover and mutation operators to generate new trial loading patterns (LPs). A core physics calculation code for fuel LP optimization of VVER reactors (LPO-V) has been developed and verified based on a VVER-1000 MOX benchmark core in comparison with MCNP4c calculations. Calculations for optimizing fuel LP of the VVER-1000 MOX core have been conducted using the ESA method in comparison with simulated annealing (SA) and adaptive simulated annealing (ASA). Statistical differences between these methods were also evaluated based on the Mann–Whitney U test. The results show that the ESA method is advantageous over the SA and ASA.

DOI： 10.1016/j.anucene.2020.107938

Compression of Cross-Section Data Size for High-Resolution Core Analysis Using Dimensionality Reduction Technique

Yamamoto M., Endo T., Yamamoto A.

Nuclear Science and Engineering 195 巻 ( 1 ) 頁： 33 - 49 2021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

Compression of cross-section data used for high-resolution core analysis is performed using a dimensionality reduction technique based on the singular value decomposition (SVD) and low-rank approximation. The size of cross-section data can be a significant issue in high-resolution core analysis using detailed energy and spatial resolutions. This study addresses this issue focusing on the similarity of multigroup cross sections among different spatial regions. A data compression method using the SVD and low-rank approximation is applied for the multigroup microscopic cross sections of heterogeneous material regions obtained by a lattice physics calculation with burnup and branch calculations. Weighting by nuclide number densities and neutron spectra is considered to improve the efficiency of compression for cross sections. Single-assembly transport calculations with the method of characteristics are carried out using the original cross sections and the reconstructed cross sections after data compression. The accuracy of data compression for cross sections is evaluated by comparing the multiplication factor and multigroup scalar fluxes. The results indicate that the present data compression for microscopic cross sections can reduce approximately 99.7% of the original cross-section data size under the present calculation condition.

DOI： 10.1080/00295639.2020.1781482

Fast reproduction of time-dependent diffusion calculations using the reduced order model based on the proper orthogonal and singular value decompositions

Tsujita K., Endo T., Yamamoto A.

Journal of Nuclear Science and Technology 58 巻 ( 2 ) 頁： 173 - 183 2021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：Journal of Nuclear Science and Technology

An efficient reduced order model (ROM) for time-dependent diffusion calculations using the proper orthogonal decomposition (POD) is proposed. Employing the singular value decomposition (SVD) and low-rank approximation (LRA) for the flux distributions sampled from the detail full order model (FOM) solutions, the orthogonal basis suitable for a target problem is numerically obtained. In the present ROM, flux distribution is expanded with an orthogonal basis on space. Then, the dimensionality reduction is performed for the neutron diffusion equation using the orthogonal basis, and the equation for the expansion coefficients is obtained. Since any flux distributions can be used to construct the orthogonal bases, different orthogonal bases calculated from different flux distribution sets are tested. The accuracy and computation time of the present ROM are verified in the TWIGL benchmark problem. The calculation results show that the present ROM is approximately 100 times faster than the FOM for kinetic calculations in the present conditions. The present method can be substituted as real-time FOM simulations when typical flux distributions of a target problem can be precalculated to represent the solution space with less degree of freedom (DOF).

DOI： 10.1080/00223131.2020.1814891

A New Resonance Calculation Method Using Energy Expansion Based on a Reduced Order Model

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

Nuclear Science and Engineering 195 巻 ( 7 ) 頁： 694 - 716 2021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 Resonance calculation using energy Spectrum Expansion (RSE) method is newly proposed in this paper. In this method, ultra-fine-group (UFG) spectra appearing in a resonance calculation are expanded by orthogonal bases on energy, which are extracted from the UFG spectra obtained in homogeneous geometry with various background cross sections using singular value decomposition and low-rank approximation. Namely, this method is based on the concept of a reduced order model. A neutron transport equation for flux moments (expansion coefficients) similar to the conventional one is derived and is numerically solved. This method applies to two benchmark problems in which a resonance interference effect and spatial self-shielding effect can appear. The results indicate that this method accurately predicts the reference effective cross sections and reaction rates obtained from direct UFG calculation in heterogeneous geometry.

DOI： 10.1080/00295639.2020.1863066

Multi-group cross section library generation by FRENDY for fast reactor neutronics calculations

Chiba G., Yamamoto A., Tada K., Endo T.

Transactions of the American Nuclear Society 124 巻 ( 1 ) 頁： 556 - 558 2021年

Perturbation-theory-based sensitivity analysis of prompt neutron decay constant for water-only system

Endo T., Noguchi A., Yamamoto A., Tada K.

Transactions of the American Nuclear Society 124 巻 ( 1 ) 頁： 184 - 187 2021年

　前のページ - 次のページ　

1
2
3
4
5
6
7
8
9
10
11
12
13
14
15
16
17
18
19
20
21
22
23