2022/08/06 更新

写真a

ヤマモト アキオ
山本 章夫
YAMAMOTO, Akio
所属
大学院工学研究科 総合エネルギー工学専攻 エネルギー安全工学 教授
大学院担当
大学院工学研究科
学部担当
工学部 エネルギー理工学科
職名
教授

学位 1

  1. エネルギー科学博士 ( 1999年4月   京都大学 ) 

研究キーワード 2

  1. 最適化などの計算科学に関する研究・中性子の高精度輸送計算法の研究・原子炉動特性方程式の解法・共鳴計算手法の研究・収束加速法の研究・不確かさ評価・原子炉安全評価

  2. ・軽水炉核設計手法の高度化に関する研究・原子炉核特性の感度解析・軽水炉燃料配置方法の最適化に関する研究・加速器駆動未臨界炉の概念設計・並列計算

研究分野 1

  1. エネルギー / 原子力工学  / 原子炉物理、原子力安全

現在の研究課題とSDGs 2

  1. 高精度炉心解析手法

  2. 原子炉の安全評価

経歴 7

  1. 名古屋大学工学研究科総合エネルギー工学専攻   教授

    2017年4月 - 現在

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    国名:日本国

  2. 名古屋大学工学研究科マテリアル理工学専攻・教授

    2010年4月 - 2017年3月

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    国名:日本国

  3. 名古屋大学工学研究科マテリアル理工学専攻・准教授

    2007年4月 - 2010年3月

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    国名:日本国

  4. 名古屋大学   室長

    2004年4月 - 2010年3月

  5. 名古屋大学工学研究科マテリアル理工学専攻・助教授

    2004年4月 - 2007年4月

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    国名:日本国

  6. 名古屋大学工学研究科原子核工学専攻・助教授

    2003年4月 - 2004年3月

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    国名:日本国

  7. 原子燃料工業株式会社

    1989年4月 - 2003年3月

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    国名:日本国

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学歴 3

  1. 京都大学   エネルギー科学研究科   エネルギー社会・環境科学専攻

    1996年4月 - 1998年3月

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    国名: 日本国

  2. 京都大学   工学研究科   原子核工学専攻

    1987年4月 - 1989年3月

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    国名: 日本国

  3. 京都大学   工学部   原子核工学科

    1983年4月 - 1987年3月

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    国名: 日本国

所属学協会 2

  1. 日本原子力学会

  2. 米国原子力学会

受賞 13

  1. 日本原子力学会賞 特賞・技術賞

    2022年3月   日本原子力学会   純国産核データ処理システムFRENDYにおける中性子多群断面積作成機能の開発

    山本章夫

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    受賞区分:国内学会・会議・シンポジウム等の賞  受賞国:日本国

  2. 日本原子力学会 計算科学技術部会 功績賞

    2021年3月   日本原子力学会 計算科学技術部会  

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    受賞区分:国内学会・会議・シンポジウム等の賞  受賞国:日本国

  3. 日本原子力学会賞 論文賞

    2019年3月   日本原子力学会   統合理論に基づく一般的な多領域形状における径方向及び方位角方向依存の共鳴自己遮蔽効果の取り扱い

    山本章夫

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    受賞区分:国内学会・会議・シンポジウム等の賞  受賞国:日本国

  4. 日本原子力学会 計算科学技術部会 業績賞

    2019年3月   日本原子力学会 計算科学技術部会  

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    受賞区分:国内学会・会議・シンポジウム等の賞  受賞国:日本国

  5. 日本原子力学会賞 論文賞

    2018年3月   日本原子力学会  

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    受賞区分:国内学会・会議・シンポジウム等の賞  受賞国:日本国

  6. 日本原子力学会賞 論文賞

    2017年3月   日本原子力学会  

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    受賞区分:国内学会・会議・シンポジウム等の賞  受賞国:日本国

  7. Fellow, American Nuclear Society

    2016年11月   American Nuclear Society  

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    受賞区分:国際学会・会議・シンポジウム等の賞  受賞国:アメリカ合衆国

  8. 日本原子力学会賞 論文賞

    2014年3月   日本原子力学会  

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    受賞区分:国内学会・会議・シンポジウム等の賞  受賞国:日本国

  9. 米国原子力学会炉物理部会最優秀論文賞

    2011年11月   米国原子力学会  

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    受賞国:アメリカ合衆国

  10. 日本原子力学会論文賞

    2010年3月   日本原子力学会  

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    受賞国:日本国

  11. 米国原子力学会炉物理部会最優秀論文賞

    2009年6月   米国原子力学会炉物理部会  

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    受賞国:アメリカ合衆国

  12. 日本原子力学会賞技術賞

    2004年3月   日本原子力学会  

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    受賞国:日本国

  13. 日本原子力学会賞論文賞

    1999年3月   日本原子力学会  

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    受賞国:日本国

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論文 414

  1. Implementation of Resonance Up-scattering Treatment in FRENDY Nuclear Data Processing System

    Yamamoto Akio, Endo Tomohiro, Chiba Go, Tada Kenichi

    NUCLEAR SCIENCE AND ENGINEERING     2022年7月

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    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: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

    Web of Science

    Scopus

  2. Improvements in Computational Efficiency for Resonance Calculation Using Energy Spectrum Expansion Method

    Kondo Ryoichi, Endo Tomohiro, Yamamoto Akio, Takeda Satoshi, Koike Hiroki, Yamaji Kazuya, Asano Koji

    NUCLEAR SCIENCE AND ENGINEERING   196 巻 ( 7 ) 頁: 769 - 791   2022年7月

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    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:Nuclear Science and Engineering  

    Improvements in computational efficiency for the Resonance calculation using energy Spectrum Expansion (RSE) method are proposed in order to increase the applicability of the method for core nuclear analyses. First, efficient treatment of the neutron source for the RSE method has been newly developed. This is a balanced approach from the viewpoints of computation time and memory size, in comparison with the other approaches mentioned in a previous study [R. KONDO et al., “A New Resonance Calculation Method Using Energy Expansion Based on a Reduced Order Model,” Nucl. Sci. Eng., 195, 694 (2021)]. Second, low-rank approximation has been applied to the RSE method considering the deficit ratio of the singular value for the orthogonal basis. Computation time was reduced by ~68% while maintaining sufficient accuracy of effective cross sections. Third, the impacts of the discretization parameters in the method of characteristics on the RSE method have been investigated, and coarser conditions of the parameters were found to be appropriate from the viewpoints of computation time and accuracy of effective cross sections. Finally, RSE calculations with these improvements have been performed for the fuel assembly geometry of a light water reactor. The computation time was reduced by ~70%, and the data size of the scattering cross-section moments was approximately 3900 times smaller in comparison with the RSE calculation without the improvements.

    DOI: 10.1080/00295639.2021.2025297

    Web of Science

    Scopus

  3. ACE-FRENDY-CBZ: a new neutronics analysis sequence using multi-group neutron transport calculations

    Chiba Go, Yamamoto Akio, Tada Kenichi

    JOURNAL OF NUCLEAR SCIENCE AND TECHNOLOGY     2022年6月

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    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:Journal of Nuclear Science and Technology  

    ABSTARCT: We propose a new neutronics analysis sequence using multi-group neutron transport calculations named ACE-FRENDY-CBZ. This sequence is free from uses of any application libraries; with the ACE files as the starting point, multi-group cross sections of media comprising a target system are calculated with the FRENDY code, and multi-group neutron transport calculations are performed with modules of the CBZ code system. The ACE-FRENDY-CBZ sequence was tested against the eight fast neutron systems, and good agreement in the neutron multiplication factors with the reference Monte Carlo results was obtained within 20 pcm differences in the bare systems and within 60 pcm differences in the reflected systems. It was also found that the adoption of the consistent P approximation increases the errors. In order to investigate this issue, we adopted the sub-group method to calculate spatially-dependent current-weighted total cross sections in the reflector regions, and it was suggested that the uses of the spatially-dependent cross sections with the consistent P approximation has a possibility to further improve the numerical accuracy.

    DOI: 10.1080/00223131.2022.2087783

    Web of Science

    Scopus

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

    Katano Ryota, Yamamoto Akio, Endo Tomohiro

    NUCLEAR SCIENCE AND ENGINEERING     2022年5月

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    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: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

    Web of Science

    Scopus

  5. Theoretical Derivation of a Unique Combination Number Hidden in the Higher-Order Neutron Correlation Factors Using the Pal-Bell Equation

    Endo Tomohiro, Nishioka Fuga, Yamamoto Akio, Watanabe Kenichi, Pyeon Cheol Ho

    NUCLEAR SCIENCE AND ENGINEERING     2022年4月

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    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:Nuclear Science and Engineering  

    The Pál-Bell equation is a backward-type master equation that describes the probability generating function (PGF) of neutron counts in a neutron multiplication system. Thanks to the Pál-Bell equation with the two-forked and the fundamental mode approximations, an analytical solution of PGF of neutron counts in a source-driven subcritical system can be theoretically derived. This theoretical derivation clarifies that the unique combination number of double factorial (2n−3)!! does exist in the ratio of the higher-order neutron correlation factors measured in a critical state even for any kind of fissile and moderator materials. Additionally, the unique combination numbers are experimentally validated for the order 3 ≤ n ≤ 6 through reactor noise measurements in actual subcritical systems. This knowledge can be used to judge whether a target system is in a deep subcritical state or to roughly estimate the magnitude of subcriticality, based on the factorial moments of the measured reactor noise in a zero-power state.

    DOI: 10.1080/00295639.2022.2049992

    Web of Science

    Scopus

  6. Impact of Angular and Spatial Source Distribution Approximations on Convergence Performance of Nonlinear Acceleration Methods for MOC in Slab Geometry

    Oshima Yoshiki, Endo Tomohiro, Yamamoto Akio

    NUCLEAR SCIENCE AND ENGINEERING   196 巻 ( 4 ) 頁: 379 - 394   2022年4月

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    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:Nuclear Science and Engineering  

    The convergence performance of nonlinear acceleration methods for the method of characteristics (MOC) with flat source (FS) approximation (FS MOC) or linear source (LS) approximation (LS MOC) is numerically investigated by focusing on the spatial and angular approximations in the acceleration calculations. The convergence of nonlinear acceleration depends on the consistency of the calculation models between the higher-order and lower-order (acceleration) methods. The convergence of four acceleration methods is evaluated to clarify the relationship between model consistency and convergence performance. These methods consist of FS or LS for the spatial source distribution and P1 or discrete angle for the angular distribution, i.e., (1) FS analytic coarse mesh finite difference (ACMFD) acceleration (FS ACMFD), (2) LS ACMFD, (3) FS angular-dependent discontinuity factor MOC (ADMOC) acceleration (FS ADMOC), and (4) LS ADMOC. The ACMFD and ADMOC accelerations are based on P1 and discrete angle approximations, respectively. The FS MOC and LS MOC are considered higher-order methods. The FS MOC and LS MOC with five acceleration methods, i.e., the aforementioned four acceleration methods and the conventional coarse mesh finite difference acceleration method, are used to perform fixed-source calculations in one-group one-dimensional homogeneous slab geometry, and the spectral radii are numerically evaluated. The numerical results indicate that (1) the nonlinear acceleration methods that are unconditionally stable for FS MOC also show similar convergence properties for LS MOC in one-dimensional slab geometry; (2) better convergence is observed when the consistency of higher- and lower-order models is high; and (3) when a coarse mesh is optically thick, the spatial homogenization degrades the convergence performance, even if spatial and angular approximations are consistent between the higher- and lower-order models.

    DOI: 10.1080/00295639.2021.1982549

    Web of Science

    Scopus

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

    Endo Tomohiro, Nishioka Fuga, Yamamoto Akio, Watanabe Kenichi, Pyeon Cheol Ho

    JOURNAL OF NUCLEAR SCIENCE AND TECHNOLOGY     2022年2月

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    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: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

    Web of Science

    Scopus

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

    Nishioka Fuga, Endo Tomohiro, Yamamoto Akio, Yamanaka Masao, Pyeon Cheol Ho

    NUCLEAR SCIENCE AND ENGINEERING   196 巻 ( 2 ) 頁: 133 - 143   2022年2月

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    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: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

    Web of Science

    Scopus

  9. 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

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    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

    CiNii Research

  10. 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

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    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  

    DOI: 10.3327/jaesjb.64.3_176

    CiNii Research

  11. Adaptive setting of background cross sections for generation of effective multi-group cross sections in FRENDY nuclear data processing code

    Yamamoto Akio, Endo Tomohiro, Tada Kenichi

    JOURNAL OF NUCLEAR SCIENCE AND TECHNOLOGY   58 巻 ( 12 ) 頁: 1343 - 1350   2021年12月

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    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:Journal of Nuclear Science and Technology  

    An adaptive setting method of background cross sections is implemented to FRENDY/MG, which is a multi-group neutron cross section generation module, for accurate interpolation of self-shielding factors with a minimum number of background cross sections. Since the dependence of self-shielding factors on background cross section is significantly different among energy group, reaction type, and nuclide, appropriate setting of background cross sections usually requires considerable works. In the present adaptive setting method, the range of background cross section is initially divided into 10 equal intervals and unnecessary background cross section points are eliminated. Then interpolation accuracy at each interval is tested. If the interpolation accuracy in an interval is not sufficient, the interval is successively halved until sufficient interpolation accuracy is obtained. For accurate interpolation of self-shielding factor or reaction rates, the monotone cubic interpolation is used. Verification calculations are carried out for all isotopes in JENDL-4.0. Calculation results indicate that the present method provides an appropriate set of background cross sections while satisfying input error tolerance for self-shielding factors or reaction rates. Typical numbers of background cross sections are from 10 to 25 when the monotone cubic interpolation and error tolerance of 0.01 for self-shielding factors are used.

    DOI: 10.1080/00223131.2021.1944930

    Web of Science

    Scopus

  12. Application of continuous Markov-chain Monte-Carlo method to multi-unit risk evaluations considering interdependence of accident progression among multiple units

    Sawada Kento, Yamamoto Akio, Endo Tomohiro, Sato Chikahiro, Maeda Keisuke, Jang Sunghyon

    JOURNAL OF NUCLEAR SCIENCE AND TECHNOLOGY   58 巻 ( 12 ) 頁: 1308 - 1317   2021年12月

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    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:Journal of Nuclear Science and Technology  

    The accident in Fukushima Dai-ichi nuclear power plants reconfirms the necessity of the safety assessment considering multiple nuclear reactor units. However, consideration of the interdependency among safety systems or events in multiple units as well as the time dependency of accident progressions is difficult in the conventional event tree method, which is widely used in probabilistic risk assessments. Recently, the continuous Markov-chain Monte-Carlo (CMMC) method coupled with a plant safety analysis or a severe accident analysis code has been paid attention to address these issues. In the present study, the CMMC coupling method is applied to the risk assessment of multiple units to clarify the benefits and issues to be resolved of this method. Since the CMMC coupling method requires many executions of an accident analysis code, a meta-model that simplifies systems and physical phenomena in accidents is used in this study to reduce computational cost. Furthermore, the inverse transform sampling method is newly adapted. Numerical analyses of BWR accident under station blackout with loss of cooling capability are carried out considering the correlation among the availabilities of mitigation systems. The results suggested that the CMMC coupling method can quantitatively treat the interdependency and time dependency among events in multiple units.

    DOI: 10.1080/00223131.2021.1940341

    Web of Science

    Scopus

  13. Multi-group neutron cross section generation capability for FRENDY nuclear data processing code

    Yamamoto Akio, Tada Kenichi, Chiba Go, Endo Tomohiro

    JOURNAL OF NUCLEAR SCIENCE AND TECHNOLOGY   58 巻 ( 11 ) 頁: 1165 - 1183   2021年11月

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    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:Journal of Nuclear Science and Technology  

    The multi-group cross-section generation capability for neutrons is implemented in the FRENDY nuclear data processing code. ACE-formatted files are used as the source of nuclear data instead of ENDF-formatted files since FRENDY already has the capability to generate pointwise cross sections in the ACE format. The several distinguished features are implemented for the multi-group generation capability, e.g. explicit consideration of resonance interference effect among nuclides, enhanced resonance treatment for various nuclear reactions, and accurate numerical integration of thermal cross sections. Verification calculations of the newly implemented capability are carried out through the comparison with the NJOY nuclear data processing code. Cross-section generations for all nuclides in JENDL-4.0, -4.0u, - (Formula presented.), ENDF/B-VII.1, -VIII.0, JEFF-3.3, and TENDL-2019 are carried out without unexpected processing issues, except for Pu-238 of TENDL-2019 that includes inconsistent data. The verification results indicate that the multi-group cross sections generated by FRENDY are consistent with those generated by NJOY. Now FRENDY can generate not only the pointwise cross sections for continuous energy Monte-Carlo codes but also the multi-group cross sections for deterministic neutronics analysis codes.

    DOI: 10.1080/00223131.2021.1921631

    Web of Science

    Scopus

  14. Proposal and applicability of estimated criticality lower-limit multiplication factor using the bootstrap method

    Hayashi Takuto, Endo Tomohiro, Yamamoto Akio

    JOURNAL OF NUCLEAR SCIENCE AND TECHNOLOGY   58 巻 ( 9 ) 頁: 1008 - 1017   2021年9月

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    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:Journal of Nuclear Science and Technology  

    To judge whether an application system is a subcritical state or not based on numerical results of the effective neutron multiplication factor (Formula presented.), an evaluation method of the estimated criticality lower-limit multiplication factor (ECLLMF) using the bootstrap method is newly proposed. By utilizing numerical results of (Formula presented.) for critical benchmark-problems that are selected depending on neutronic similarity to the application system, the ECLLMF should be carefully and conservatively estimated based on uncertainties of (Formula presented.) due to a criticality safety analysis code, experimental uncertainty, and covariance matrix of the nuclear data library. Furthermore, the frequency distribution of (Formula presented.) for these problems does not necessarily obey an ideal normal distribution. Using a resampling technique called ‘bootstrap method,’ the proposed method can reasonably estimate the ECLLMF considering the uncertainties and nuclear data-induced correlation between each critical benchmark-problem without the assumption of normality. To investigate the applicability of the proposed method, the approach-to-criticality experiment was carried out at the Kyoto University Critical Assembly (KUCA). Comparison of numerical results of (Formula presented.) and the ECLLMF using the bootstrap method indicated that the proposed method was able to judge an actual subcritical core as subcritical state.

    DOI: 10.1080/00223131.2021.1902416

    Web of Science

    Scopus

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

    Kondo Ryoichi, Endo Tomohiro, Yamamoto Akio, Takeda Satoshi, Koike Hiroki, Yamaji Kazuya, Sato Daisuke

    NUCLEAR SCIENCE AND ENGINEERING   195 巻 ( 7 ) 頁: 694 - 716   2021年7月

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    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: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

    Web of Science

    Scopus

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

    Tsujita Kosuke, Endo Tomohiro, Yamamoto Akio

    JOURNAL OF NUCLEAR SCIENCE AND TECHNOLOGY   58 巻 ( 2 ) 頁: 173 - 183   2021年2月

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    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: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

    Web of Science

    Scopus

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

    Viet-Phu Tran, Phan Giang T. T., Van-Khanh Hoang, Pham Nhu Viet Ha, Yamamoto Akio, Hoai-Nam Tran

    ANNALS OF NUCLEAR ENERGY   151 巻   2021年2月

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    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: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

    Web of Science

    Scopus

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

    Sasaki Miyuki, Sanada Yukihisa, Katengeza Estiner W., Yamamoto Akio

    SCIENTIFIC REPORTS   11 巻 ( 1 )   2021年1月

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    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: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

    Web of Science

    Scopus

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

    Yamamoto Masato, Endo Tomohiro, Yamamoto Akio

    NUCLEAR SCIENCE AND ENGINEERING   195 巻 ( 1 ) 頁: 33 - 49   2021年1月

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    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: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

    Web of Science

    Scopus

  20. Verification of the multi-group generation capability of FRENDY nuclear data processing code for recent nuclear data through comparison of one-group reaction rates

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

    Transactions of the American Nuclear Society   124 巻 ( 1 ) 頁: 544 - 547   2021年

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    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:Transactions of the American Nuclear Society  

    DOI: 10.13182/T124-35126

    Scopus

  21. 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

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    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:Transactions of the American Nuclear Society  

    DOI: 10.13182/T124-35116

    Scopus

  22. 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

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    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:Transactions of the American Nuclear Society  

    DOI: 10.13182/T124-35323

    Scopus

  23. Data Assimilation Using Subcritical Measurement of Prompt Neutron Decay Constant

    Endo Tomohiro, Yamamoto Akio

    NUCLEAR SCIENCE AND ENGINEERING   194 巻 ( 11 ) 頁: 1089 - 1104   2020年11月

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    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:Nuclear Science and Engineering  

    The prompt neutron decay constant (Formula presented.) in a steady-state subcritical system can be directly measured using a reactor noise analysis method such as the Feynman- (Formula presented.) method. To reduce the nuclear data–induced uncertainty of (Formula presented.) for a target system, this study investigates the applicability of data assimilation techniques, i.e., the bias factor method and the cross-section adjustment method, based on a subcritical measurement of (Formula presented.) conducted at Kyoto University Critical Assembly (KUCA). The sensitivity coefficients of (Formula presented.) and (Formula presented.) with respect to the nuclear data were efficiently estimated using a deterministic SN transport code with first-order perturbation theory. As a result, the a priori relative uncertainty of (Formula presented.) due to the 56-group SCALE covariance data can be reduced if there is strong correlation between the measured (Formula presented.) and the target (Formula presented.). The experimental value of (Formula presented.) contributes to improving the nuclear data of total fission spectrum (Formula presented.) and total fission neutron number (Formula presented.) via strong correlations between (Formula presented.) and prompt (Formula presented.) and between (Formula presented.) and prompt (Formula presented.), by utilizing the sensitivity coefficients of (Formula presented.) with respect to prompt (Formula presented.) and (Formula presented.).

    DOI: 10.1080/00295639.2020.1720499

    Web of Science

    Scopus

  24. Implementation of the unscented transformation with low rank approximation in uncertainty analysis during large-break loss of coolant accident 国際共著

    Foad Basma, Yamamoto Akio, Endo Tomohiro

    ANNALS OF NUCLEAR ENERGY   146 巻   2020年10月

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    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:Annals of Nuclear Energy  

    The Low Rank Approximation (LRA) and Unscented Transform (UT) are integrated to produce a new algorithm having the capability to decrease the time required for the uncertainty quantification during Loss of coolant accident (LOCA) in Pressurized Water Reactors (PWR). The LRA is an efficient technique used in reducing computational cost due to its ability to perform dimensionality reduction by revealing the active or important degrees of freedom and calculate the basis of the so-called active subspace basing on the Singular Value Decomposition (SVD). For further reduction in the computational time; the UT algorithm is also implemented to generate a set of sigma points, these sigma points are the representatives of the whole probability distribution (the UT is restricted to Gaussian distribution). The main safety parameter is the maximum cladding temperatures during the accident which are computed by ATHLET thermal-hydraulic code. The reactivity coefficients and the covariance matrix are calculated using the SCALE 6.2 code. The present calculation model has 14-dimensions, therefore the number of sigma points needed for the SVD/UT technique is 29, and can be minimized to 5 sigma points only if the LRA/UT is used where two singular values are sufficient to reproduce/span the space thanks to the strong correlations between the reactivity coefficients.

    DOI: 10.1016/j.anucene.2020.107614

    Web of Science

    Scopus

  25. Uncertainty and regression analysis of the MSLB accident in PWR based on unscented transformation and low rank approximation 国際共著

    Foad Basma, Yamamoto Akio, Endo Tomohiro

    ANNALS OF NUCLEAR ENERGY   143 巻   2020年8月

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    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:Annals of Nuclear Energy  

    The present studies focus on the quantification of uncertainty during the main steam line break accident scenario (MSLB) in PWR, assuming that there is a failure on the feed-water regulating valve of the broken steam generator. The scenario is characterized by the associated positive Doppler and coolant density reactivities which bring the core back to critical (return-to-power). Accordingly, the input uncertainty parameters are the Doppler and coolant density reactivities taking into account the correlation matrix among the input parameters, which is calculated by SCALE 6.2 code. The main safety parameters are the maximum cladding surface temperatures and average core power during the accident which are computed by ATHLET thermal-hydraulic code. The sampling-based uncertainty technique is considered to be the most dependable technique which can be applicable to any code, however it is computationally expensive. Therefore, it is important to develop efficient techniques which are capable of reducing the calculation time. The first approach is the SVD-UT where the Unscented Transform (UT) algorithm and singular value decomposition (SVD) are combined to generate a minimal sample points. In addition, due to the strong correlation between the input reactivities, the computational time can be further reduced by implementing the Low Rank Approximation (LRA) and revealing the active subspace.

    DOI: 10.1016/j.anucene.2020.107493

    Web of Science

    Scopus

  26. Applicability of a reduced order model for a safety analysis code to statistical safety analysis 査読有り

    Matsushita Masaki, Endo Tomohiro, Yamamoto Akio

    JOURNAL OF NUCLEAR SCIENCE AND TECHNOLOGY     頁: 1-12   2020年7月

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    記述言語:英語   掲載種別:研究論文(学術雑誌)  

    DOI: 10.1080/00223131.2020.1783382

    Web of Science

    Scopus

  27. Efficient uncertainty quantification for PWR during LOCA using unscented transform with singular value decomposition 国際共著

    Foad Basma, Yamamoto Akio, Endo Tomohiro

    ANNALS OF NUCLEAR ENERGY   141 巻   2020年6月

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    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:Annals of Nuclear Energy  

    This paper discusses one of the most important issues facing the regulatory body while performing the uncertainty analysis of the nuclear reactor parameter during accident conditions. This problem is the long computational time required by the statistical sampling methods to compute the uncertainty. We overcome this problem by introducing the Unscented Transform (UT) algorithm and singular value decomposition (SVD). Where both algorithms are combined (SVD/UT) to generate a set of sigma points, these sigma points are the representatives of whole probability distribution. The uncertainty quantification is performed during Loss of coolant accident in Pressurized Water Reactor (PWR), where the input variable of uncertainty is the coolant density reactivity. The SCALE 6.2 code is used for calculating the reactivity coefficients and the covariance matrix. The response variables are the peak cladding temperatures during the accident which are computed by ATHLET thermal-hydraulic code. The results obviously confirm the efficiency of the SVD/UT sampling in predicting the new mean values, and assure its ability to reduce the sampling size leading to a dramatic reduction of computational cost.

    DOI: 10.1016/j.anucene.2020.107341

    Web of Science

    Scopus

  28. Impact of Various Parameters on Convergence Performance of CMFD Acceleration for MOC in Multigroup Heterogeneous Geometry

    Oshima Yoshiki, Endo Tomohiro, Yamamoto Akio, Kodama Yasuhiro, Ohoka Yasunori, Nagano Hiroaki

    NUCLEAR SCIENCE AND ENGINEERING   194 巻 ( 6 ) 頁: 477 - 491   2020年6月

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    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:Nuclear Science and Engineering  

    The impact of various parameters in the coarse mesh finite difference (CMFD) acceleration method on overall convergence behavior is investigated through numerical calculations using the method of characteristics (MOC). Four parameters appearing in the CMFD acceleration with MOC, i.e., scalar flux distribution in flat flux regions (FFRFlux), the scalar flux distribution in CMFD meshes (CMFDFlux), homogenized cross sections (HXSs) in CMFD meshes, and current correction factors (CCFs), are considered. Parts of these four parameters are fixed to the converged values throughout iterations in order to estimate their impact on convergence. Numerical calculations are carried out for Korea Advanced Institute of Science and Technology’s (KAIST’s) benchmark problem KAIST-2A, which is a heterogeneous and multigroup problem, and the number of outer iterations to reach convergence is evaluated. The impact of geometric heterogeneity and cross-section homogenization in the CMFD acceleration has not been considered in linearized Fourier analysis so far. The calculation results indicate that (1) convergence of HXS has little impact on the overall convergence, (2) convergence of FFRFlux is dominant followed by CCF when a CMFD mesh is optically thin, and (3) convergence of FFRFlux is dominant when a CMFD mesh is optically thick and contains many flat flux regions.

    DOI: 10.1080/00295639.2020.1722512

    Web of Science

    Scopus

  29. Application of the multigrid amplitude function method for time-dependent MOC based on the linear source approximation 査読有り

    Tsujita Kosuke, Endo Tomohiro, Yamamoto Akio

    JOURNAL OF NUCLEAR SCIENCE AND TECHNOLOGY     2020年1月

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    記述言語:英語   掲載種別:研究論文(学術雑誌)  

    DOI: 10.1080/00223131.2019.1709993

    Web of Science

  30. Subcriticality measurement using time-domain decomposition-based integral method for simultaneous reactivity and source changes 査読有り

    Endo Tomohiro, Nonaka Asahi, Imai Sho, Yamamoto Akio, Sakon Atsushi, Hashimoto Kengo

    JOURNAL OF NUCLEAR SCIENCE AND TECHNOLOGY     2020年1月

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    記述言語:英語   掲載種別:研究論文(学術雑誌)  

    DOI: 10.1080/00223131.2019.1706658

    Web of Science

  31. A kinetic model for the direct response matrix method

    Mitsuyasu Takeshi, Yamamoto Akio

    JOURNAL OF NUCLEAR SCIENCE AND TECHNOLOGY   57 巻 ( 1 ) 頁: 90 - 99   2020年1月

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    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:Journal of Nuclear Science and Technology  

    An accurate analysis model for transient reactor behavior is necessary to keep sufficient safety margins of nuclear power plants to prevent cliff edge effects. In this study, the direct response matrix (DRM) method is applied to the kinetic domain and the transient analysis is enabled based on the transport equation. The kinetic DRM model introduces the time delay to four sub-response matrices. The time delay can be evaluated by a Monte Carlo calculation. The model is evaluated in homogeneous and heterogeneous problems. The Doppler feedback is considered in the heterogeneous problem and the calculation results are compared with the experimental data. The calculation results indicate that the calculation step 1.0E-7 s is sufficient for the model and the model provides results in good agreement with the experimental data. It is concluded that the present model with the DRM method can be used for transient analysis.

    DOI: 10.1080/00223131.2019.1659874

    Web of Science

    Scopus

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

    Tsujita K., Endo T., Yamamoto A.

    Transactions of the American Nuclear Society   123 巻 ( 1 ) 頁: 1349 - 1353   2020年

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    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:Transactions of the American Nuclear Society  

    DOI: 10.13182/T123-33370

    Scopus

  33. Subcriticality estimation using unscented Kalman filter for reactivity- And source-transients

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

    Transactions of the American Nuclear Society   123 巻 ( 1 ) 頁: 841 - 844   2020年

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    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:Transactions of the American Nuclear Society  

    DOI: 10.13182/T123-33367

    Scopus

  34. 計算科学を活用した炉物理研究の最先端

    山本 章夫

    年次大会   2020 巻 ( 0 ) 頁: K08101   2020年

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    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 日本機械学会  

    DOI: 10.1299/jsmemecj.2020.k08101

    CiNii Research

  35. A resonance calculation method using energy expansion based on a reduced order model: Use of ultra-fine group spectrum calculation and application to heterogeneous geometry

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

    International Conference on Physics of Reactors: Transition to a Scalable Nuclear Future, PHYSOR 2020   2020-March 巻   頁: 152 - 161   2020年

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    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 Physics of Reactors: Transition to a Scalable Nuclear Future, PHYSOR 2020  

    A Resonance calculation using energy Spectral Expansion (RSE) method has been recently proposed in order to efficiently treat complicated heterogeneous geometry and resonance interference effect. In the RSE method, ultra-fine group spectra are generated from ultra-fine group calculations in homogeneous geometry, and the spectra are expanded by the orthogonal basis on energy based on the singular value decomposition. Then the transport calculation for expansion coefficients is numerically performed, and the ultra-fine group spectra in the target heterogeneous regions are reconstructed by the expansion coefficients and the orthogonal basis. In this study, the RSE method is applied to multi-cell geometries including UO2, MOX and water cells, in which the resonance interference effect between UO2 and MOX fuel cells appears. The validity of the RSE method is confirmed through comparison with the reference effective multi-group cross sections obtained from the direct ultra-fine group calculation in the target heterogeneous geometry.

    DOI: 10.1051/epjconf/202124702006

    Scopus

  36. Experiment of unique combination number due to the third-order neutron-correlation

    Endo T., Imai S., Watanabe K., Yamamoto A., Sakon A., Hashimoto K., Yamanaka M., Pyeon C.H.

    International Conference on Physics of Reactors: Transition to a Scalable Nuclear Future, PHYSOR 2020   2020-March 巻   頁: 1736 - 1744   2020年

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    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 Physics of Reactors: Transition to a Scalable Nuclear Future, PHYSOR 2020  

    From zero-power reactor noise measurement, the second- and third-order neutron correlation factors Y and y3 can be evaluated by analyzing mean, variance, the third-order central moment of neutron count data. Theoretically, it is expected that the neutron-correlation ratio Y3/Y2 converges to the unique combination number “3” at a near-critical state in an arbitrary system without depending on the fissile material and the neutron-energy spectrum of core, as the neutron counting gate width T increases sufficiently. Thus, the information about the difference between y3/Y2 and “3” has the potential to judge whether a target unknown system is critical or not and to roughly guess the absolute value of subcriticality. In this study, the detector dead-time effect on y3/Y2 is theoretically investigated based on the heuristic method using the single-, pair-, and trio-detection probabilities with the fundamental mode approximation. As a result, it is clarified that the saturation value of y3/Y2 converges to “3” independent of the dead time, when a target system is a critical state. For validation, actual experimental results are presented for a non-multiplication system driven by 252Cf spontaneous source, and shallow and deep subcritical systems at Japanese experimental facilities (UTR-KINKI and KUCA) under the shutdown state. Consequently, it is demonstrated that y3/Y2 shows a significant difference from “3” in the non-multiplication system. In the case of subcritical systems driven by inherent neutron sources, it is confirmed that the ratios y3/Y2 are close to the unique combination number “3,” and the slight difference from “3” is measurable by the long-time reactor noise measurement for the deep subcritical system.

    DOI: 10.1051/epjconf/202124709004

    Scopus

  37. Estimated criticality lower-limit multiplication factor of low-enriched uranium dioxide-concrete system using the bootstrap method

    Hayashi T., Nishioka F., Endo T., Yamamoto A.

    International Conference on Physics of Reactors: Transition to a Scalable Nuclear Future, PHYSOR 2020   2020-March 巻   頁: 2690 - 2697   2020年

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    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 Physics of Reactors: Transition to a Scalable Nuclear Future, PHYSOR 2020  

    The present paper aims to evaluate the estimated criticality lower-limit multiplication factor of fuel debris in a form of uranium dioxide-concrete mixture for a study of criticality control on the fuel debris generated through the molten core concrete interaction in a severe accident of a light water reactor. The estimated criticality lower-limit multiplication factor is evaluated using the bootstrap method where the assumption of the normal distribution is not necessary. In addition, it is calculated taking into account correlation coefficients that represent the degrees of neutronic similarity between the target system and benchmark critical experiment systems, experimental uncertainties of benchmark data, and statistical uncertainties of calculated effective multiplication factor by a continuous energy Monte Carlo code. This paper shows that the estimated criticality lower-limit multiplication factor using the bootstrap method can be comparable with a baseline upper-subcritical-limit which is evaluated by Whisper-1.1 without margins of subcriticality for uncertainties from nuclear covariance data and undetected errors in software.

    DOI: 10.1051/epjconf/202124717001

    Scopus

  38. Development of FRENDY nuclear data processing code: Generation capability of multi-group cross sections from ace file

    Yamamoto A., Endo T., Tada K.

    Transactions of the American Nuclear Society   122 巻   頁: 714 - 717   2020年

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    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:Transactions of the American Nuclear Society  

    DOI: 10.13182/T122-32047

    Scopus

  39. Application of bootstrap method to estimated criticality lower-limit multiplication factor considering nuclear data-induced uncertainty

    Hayashi T., Endo T., Yamamoto A.

    Transactions of the American Nuclear Society   122 巻   頁: 458 - 461   2020年

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    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:Transactions of the American Nuclear Society  

    DOI: 10.13182/T122-32039

    Scopus

  40. Loading pattern optimization for a PWR using Multi-Swarm Moth Flame Optimization Method with Predator 査読有り

    Ishiguro Satomi, Endo Tomohiro, Yamamoto Akio

    JOURNAL OF NUCLEAR SCIENCE AND TECHNOLOGY     2019年12月

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    記述言語:英語   掲載種別:研究論文(学術雑誌)  

    DOI: 10.1080/00223131.2019.1700844

    Web of Science

  41. Development of a reduced order model for severe accident analysis codes by singular value decomposition aiming probabilistic safety margin analysis 査読有り

    Matsushita Masaki, Endo Tomohiro, Yamamoto Akio

    JOURNAL OF NUCLEAR SCIENCE AND TECHNOLOGY     2019年12月

     詳細を見る

    記述言語:英語   掲載種別:研究論文(学術雑誌)  

    DOI: 10.1080/00223131.2019.1699190

    Web of Science

  42. 破損・堆積現象の解明を目指して(<特集>廃炉国際ワークショップ FDR2019)

    倉田 正輝, 山本 章夫, 溝上 伸也

    日本機械学会誌   122 巻 ( 1211 ) 頁: 10 - 12   2019年10月

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    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 日本機械学会  

    DOI: 10.1299/jsmemag.122.1211_10

    CiNii Research

  43. APPLICATION OF THE FOREST SHIELDING FACTOR TO THE MAXIMUM-LIKELIHOOD EXPECTATION MAXIMIZATION METHOD FOR AIRBORNE RADIATION MONITORING 査読有り

    Sasaki M., Sanada Y., Yamamoto A.

    RADIATION PROTECTION DOSIMETRY   184 巻 ( 3-4 ) 頁: 400-404   2019年10月

     詳細を見る

    記述言語:英語   掲載種別:研究論文(学術雑誌)  

    DOI: 10.1093/rpd/ncz095

    Web of Science

  44. Transport consistent diffusion coefficient for CMFD acceleration and comparison of convergence properties

    Yamamoto Akio, Endo Tomohiro, Giho Akinori

    JOURNAL OF NUCLEAR SCIENCE AND TECHNOLOGY   56 巻 ( 8 ) 頁: 716 - 723   2019年8月

     詳細を見る

    記述言語:日本語   掲載種別:研究論文(学術雑誌)  

    DOI: 10.1080/00223131.2019.1618405

    Web of Science

  45. A simple treatment of increased gap due to fuel assembly bowing through correction of cross sections

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

    JOURNAL OF NUCLEAR SCIENCE AND TECHNOLOGY   56 巻 ( 6 ) 頁: 471 - 478   2019年6月

     詳細を見る

    記述言語:日本語   掲載種別:研究論文(学術雑誌)  

    DOI: 10.1080/00223131.2019.1598509

    Web of Science

  46. Development of assembly bowing model for pin-by-pin core calculations

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

    International Conference on Nuclear Engineering, Proceedings, ICONE   2019-May 巻   2019年5月

     詳細を見る

    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 Nuclear Engineering, Proceedings, ICONE  

    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.

    Scopus

  47. Experimental validation of unique combination numbers for third- and fourth-order neutron correlation factors of zero-power reactor noise

    Endo Tomohiro, Yamamoto Akio, Yamanaka Masao, Pyeon Cheol Ho

    JOURNAL OF NUCLEAR SCIENCE AND TECHNOLOGY   56 巻 ( 4 ) 頁: 322 - 336   2019年4月

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    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:Journal of Nuclear Science and Technology  

    Zero-power reactor noise is useful for subcriticality measurements. Based on the nuclear reactor physics and the theory of neutron detection, this paper theoretically clarifies that the third- and fourth-order neutron correlation factors Y 3 and Y 4 can be expressed as functions of the second-order neutron correlation factor Y. In particular, if the neutron-counting gate width is sufficiently large, the saturation values Y 3 /Y 2 and Y 4 /Y 2 are almost equal to the unique combination numbers, ‘3’ and ‘15,’ for a source-driven subcritical system, where the subcriticality is less than 10,000 pcm. These unique combination numbers, ‘3’ and ‘15,’ for Y 3 /Y 2 and Y 4 /Y 3 were validated using actual zero-power reactor noise measurements carried out at the Kyoto University Criticality Assembly. In this study, the estimation of statistical errors and correlations between different gate widths owing to the bunching method was achieved by the moving block bootstrap method. For a sufficiently long measured reactor noise in a steady and unperturbed state, a statistical test for the evaluation of the critical state and the absolute measurement of subcriticality can be carried out by statistically quantifying the difference between the measurement value of Y 3 /Y 2 and the unique combination number.

    DOI: 10.1080/00223131.2019.1580625

    Web of Science

    Scopus

  48. Comparison of theoretical formulae and bootstrap method for statistical error estimation of Feynman-alpha method

    Endo Tomohiro, Yamamoto Akio

    ANNALS OF NUCLEAR ENERGY   124 巻   頁: 606 - 615   2019年2月

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    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: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

  49. Inverse estimation methods of unknown radioactive source for fuel debris search

    Sugaya Shinji, Endo Tomohiro, Yamamoto Akio

    ANNALS OF NUCLEAR ENERGY   124 巻   頁: 49 - 57   2019年2月

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    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: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

  50. A New Interpretation of Discontinuity Factor

    Yamamoto Akio, Endo Tomohiro

    NUCLEAR SCIENCE AND ENGINEERING   193 巻 ( 9 ) 頁: 991 - 997   2019年

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    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: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

  51. Utilization of Regionwise Even-Parity Discontinuity Factor to Reduce Discretization Error of MOC

    Yamamoto Akio, Giho Akinori, Endo Tomohiro

    NUCLEAR SCIENCE AND ENGINEERING   193 巻 ( 3 ) 頁: 253 - 268   2019年

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    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:Nuclear Science and Engineering  

    To reduce angular and spatial discretization error of the method of characteristics with a coarse calculation condition, the regionwise even-parity discontinuity factor (EPDF) for transport calculations is evaluated through an iterative procedure using only the regionwise scalar flux, i.e., without the odd-parity angular flux, the partial current, or the net current at the region boundary. The regionwise EPDF is evaluated in a single-assembly geometry with the reflective boundary condition. The evaluated EPDF is applied to a 2 × 2 colorset assembly and core configurations, and the performance is compared to that of the conventional superhomogenization (SPH) method. The calculation results indicate that (1) no convergence issue is observed during the iteration process to estimate the EPDF, (2) the performance of the regionwise EPDF is better than that of the conventional SPH method, and (3) the normalization of the EPDF is necessary to incorporate different surface scalar flux levels among different types of fuel assemblies.

    DOI: 10.1080/00295639.2018.1516961

    Web of Science

    Scopus

  52. 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

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    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

  53. 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

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    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

  54. 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

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    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  

    DOI: 10.3327/jaesjb.61.4_254

    CiNii Research

  55. 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

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    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:Transactions of the American Nuclear Society  

    Scopus

  56. 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

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    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:Transactions of the American Nuclear Society  

    DOI: 10.13182/T30783

    Scopus

  57. 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

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    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:Transactions of the American Nuclear Society  

    DOI: 10.13182/T30993

    Scopus

  58. 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

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    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

  59. 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

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    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  

    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

  60. 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

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    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  

    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

  61. 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

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    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 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

  62. 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

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    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

  63. 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

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    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:Transactions of the American Nuclear Society  

    DOI: 10.13182/T30744

    Scopus

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

    Hayashi Koji, Endo Tomohiro, Yamamoto Akio

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

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    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: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

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

    Mitsuyasu Takeshi, Aoyama Motoo, Yamamoto Akio

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

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    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: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

  66. Sensitivity analysis of prompt neutron decay constant using perturbation theory

    Endo Tomohiro, Yamamoto Akio

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

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    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: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

  67. 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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    記述言語:英語  

  68. 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

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    記述言語:英語  

  69. 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

     詳細を見る

    担当区分:筆頭著者   記述言語:英語  

  70. 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

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    担当区分:筆頭著者   記述言語:英語   掲載種別:研究論文(学術雑誌)  

  71. 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

     詳細を見る

    記述言語:英語  

  72. 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

     詳細を見る

    記述言語:英語  

  73. 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

     詳細を見る

    記述言語:英語  

  74. 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

     詳細を見る

    記述言語:英語  

  75. 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

     詳細を見る

    記述言語:英語  

  76. 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

     詳細を見る

    担当区分:筆頭著者   記述言語:英語  

  77. 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

     詳細を見る

    担当区分:筆頭著者   記述言語:英語  

  78. 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

     詳細を見る

    記述言語:英語   掲載種別:研究論文(学術雑誌)  

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

        頁: [USB-DRIVE]   2018年4月

     詳細を見る

    記述言語:英語  

  80. Estimation of sensitivity coefficient based on lasso-type penalized linear regression

    Katano Ryota, Endo Tomohiro, Yamamoto Akio, Tsujimoto Kazufumi

    JOURNAL OF NUCLEAR SCIENCE AND TECHNOLOGY   55 巻 ( 10 ) 頁: 1099 - 1109   2018年

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    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:Journal of Nuclear Science and Technology  

    We proposed the penalized regression ‘adaptive smooth-lasso’ for the estimation of sensitivity coefficients of the neutronics parameters. The proposed method utilizes the variation of the microscopic cross-sections and the neutronics parameters obtained by random sampling. The weighted penalty term of the proposed method is more appropriate for the estimation of the sensitivity of neutronics parameters to the microscopic cross-section than that of the conventional methods. In a numerical verification calculation, sensitivity coefficients of keff of an accelerator-driven system are estimated using the proposed method, the conventional penalized regression, and the direct method. Comparison of these results indicates that the proposed method is superior to the conventional penalized linear regression from the viewpoint of reproduction of the reference sensitivity coefficients obtained by the direct method. Through the verification calculations, the proposed method can be a candidate for a practical method to estimate the sensitivity coefficients with low calculation cost.

    DOI: 10.1080/00223131.2018.1479988

    Web of Science

    Scopus

  81. Radially and azimuthally dependent resonance self-shielding treatment for general multi-region geometry based on a unified theory

    Koike Hiroki, Kirimura Kazuki, Yamaji Kazuya, Kosaka Shinya, Yamamoto Akio

    JOURNAL OF NUCLEAR SCIENCE AND TECHNOLOGY   55 巻 ( 1 ) 頁: 41 - 65   2018年

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    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:Journal of Nuclear Science and Technology  

    A unified resonance self-shielding method, which can treat general sub-divided fuel regions, is developed for lattice physics calculations in reactor physics field. In a past study, a hybrid resonance treatment has been developed by theoretically integrating equivalence theory and ultra-fine-group slowing-down calculation. It can be applied to a wide range of neutron spectrum conditions including low moderator density ranges in severe accident states, as long as each fuel region is not sub-divided. In order to extend the method for radially and azimuthally sub-divided multi-region geometry, a new resonance treatment is established by incorporating the essence of sub-group method. The present method is composed of two-step flux calculation, i.e. ‘coarse geometry + fine energy’ (first step) and ‘fine geometry + coarse energy’ (second step) calculations. The first step corresponds to a hybrid model of the equivalence theory and the ultra-fine-group calculation, and the second step corresponds to the sub-group method. From the verification results, effective cross-sections by the new method show good agreement with the continuous energy Monte-Carlo results for various multi-region geometries including non-uniform fuel compositions and temperature distributions. The present method can accurately generate effective cross-sections with short computation time in general lattice physics calculations.

    DOI: 10.1080/00223131.2017.1384704

    Web of Science

    Scopus

  82. Dimension-reduced cross-section adjustment method based on minimum variance unbiased estimation

    Yokoyama Kenji, Yamamoto Akio, Kitada Takanori

    JOURNAL OF NUCLEAR SCIENCE AND TECHNOLOGY   55 巻 ( 3 ) 頁: 319 - 334   2018年

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    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:Journal of Nuclear Science and Technology  

    A new formulation of the cross-section adjustment methodology with the dimensionality-reduction technique has been derived in the light of the fact that it is often used under the condition of ill-posed problem, where the number of integral experimental quantities is less than the number of adjusted nuclear data parameters. This new formulation is proposed as the dimension-reduced conventional cross-section adjustment method (DRCA). The derivation of DRCA is based on the minimum variance unbiased estimation (MVUE), and the assumption of normal distribution is not used. The result of DRCA depends on a user-defined matrix that determines the dimension-reduced feature subspace. We examined three variations of DRCA, namely, DRCA1, DRCA2, and DRCA3, which employ (1) the nuclear data covariance matrix as the user-defined matrix, (2) the sensitivity coefficient matrix postmultiplied by the nuclear data covariance matrix, and (3) the sensitivity coefficient matrix, respectively. Mathematical investigation and numerical verification revealed that DRCA2 is equivalent to the currently widely used cross-section adjustment method. Moreover, DRCA3 is found to be identical to the cross-section adjustment method based on MVUE, which has been proposed in the previous study.

    DOI: 10.1080/00223131.2017.1397563

    Web of Science

    Scopus

  83. Estimation of Sensitivity Coefficient based on Lasso-type Penalized Linear Regression 査読有り

    R. Katano, T. Endo, A. Yamamoto, K. Tsujimoto

    JOURNAL OF NUCLEAR SCIENCE AND TECHNOLOGY   55 巻   頁: 1099-1109   2018年

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    記述言語:英語   掲載種別:研究論文(学術雑誌)  

  84. Utilization of Region-wise Even-Parity Discontinuity Factor to Reduce Discretization Error of MOC 査読有り

    A. Yamamoto, A. Giho, T. Endo

    Nuclear Science and Engineering   193 巻   頁: 253-268   2018年

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    担当区分:筆頭著者   記述言語:英語   掲載種別:研究論文(学術雑誌)  

  85. Flux Region Assignment Method using Ray Trace Information for the Method of Characteristics to Improve Cache Efficiency

    A. Yamamoto, A. Giho, T. Endo

    Nuclear Science and Engineering   192 巻   頁: 243-250   2018年

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    担当区分:筆頭著者   記述言語:英語   掲載種別:研究論文(学術雑誌)  

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

    Yamamoto Akio, Giho Akinori, Endo Tomohiro

    NUCLEAR SCIENCE AND ENGINEERING   192 巻 ( 3 ) 頁: 240 - 253   2018年

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    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: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

  87. リスク評価とマネジメントに関するアジアシンポジウム開催報告

    山口 彰, 山本 章夫, 成宮 祥介

    日本原子力学会誌ATOMOΣ   60 巻 ( 6 ) 頁: 362 - 363   2018年

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    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> リスクに関する国際シンポジウムは,日中韓を中心に横浜で開催された。20年ほど前に日韓で始まった日韓PSAワークショップにその起源をおくもので,原子力施設に関するリスク評価の専門家が一堂に会し,充実した意見交換,議論が行われた。リスクの本格活用に臨む我が国としては多くの知見を得たシンポジウムとなった。</p>

    DOI: 10.3327/jaesjb.60.6_362

    CiNii Research

  88. 原子炉物理分野の研究開発ロードマップ2017:次世代が考える炉物理の未来

    山本 章夫, 千葉 豪, 桐村 一生, 三木 陽介, 横山 賢治

    日本原子力学会誌ATOMOΣ   60 巻 ( 4 ) 頁: 241-245   2018年

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    記述言語:英語   掲載種別:研究論文(学術雑誌)  

    <p> 炉物理部会の傘下に設置された「炉物理ロードマップ調査・検討」WGにおけるロードマップ策定の概要を紹介する。本ロードマップの特徴は,①次世代を担う若手の技術者・研究者を中心に議論・策定を進めたこと,②現状から類推して課題を設定するフォアキャストアプローチに加え,原子炉物理分野のビジョンとミッションを検討し,これらを達成するために解決すべき課題をバックキャストアプローチにより設定したこと,にある。本ロードマップの詳細は,報告書として炉物理部会のホームページより閲覧可能である。</p>

    DOI: 10.3327/jaesjb.60.4_241

  89. 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   119 巻   頁: 900 - 903   2018年

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    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:Transactions of the American Nuclear Society  

    Scopus

  90. Quantification of modeling approximation error of pin-cell calculation using kriging and principal component analysis

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

    Transactions of the American Nuclear Society   118 巻   頁: 875 - 878   2018年

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    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:Transactions of the American Nuclear Society  

    Scopus

  91. Quantification of modeling approximation error of pin-cell calculation using kriging and principal component analysis

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

    AISTech - Iron and Steel Technology Conference Proceedings   2018-May 巻   頁: 875 - 878   2018年

     詳細を見る

    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:AISTech - Iron and Steel Technology Conference Proceedings  

    Scopus

  92. Reduction of cross section table size for core analysis using dimensionality reduction technique

    Yamamoto M., Endo T., Yamamoto A.

    Transactions of the American Nuclear Society   119 巻   頁: 1226 - 1228   2018年

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    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:Transactions of the American Nuclear Society  

    Scopus

  93. Sensitivity coefficient analysis of omega-eigenvalue based on first-order perturbation theory

    Endo T., Yamamoto A.

    International Conference on Physics of Reactors, PHYSOR 2018: Reactor Physics Paving the Way Towards More Efficient Systems   Part F168384-2 巻   頁: 1240 - 1253   2018年

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    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 Physics of Reactors, PHYSOR 2018: Reactor Physics Paving the Way Towards More Efficient Systems  

    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 ω -eigenvalue. For this purpose, the numerical calculation method using the standard first-order perturbation theory is investigated. A specific theoretical formula is derived to evaluate the sensitivity coefficient of ω to nuclear data. The derived formula utilizes forward and adjoint eigenfunctions which consist of neutron flux and delayed neutron precursor densities. Through a feasibility study based on the multi-group diffusion calculation, the derived formula is verified by comparing with the reference results using the direct method. In addition, it is confirmed that the prompt approximation is applicable to the evaluation of sensitivity coefficient of α for a subcritical state where α is sufficiently larger than decay constants of delayed neutron precursors.

    Scopus

  94. Inverse estimation methods of unknown radioactive source for fuel debris search

    Sugaya S., Endo T., Yamamoto A.

    International Conference on Physics of Reactors, PHYSOR 2018: Reactor Physics Paving the Way Towards More Efficient Systems   Part F168384-4 巻   頁: 2632 - 2643   2018年

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    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 Physics of Reactors, PHYSOR 2018: Reactor Physics Paving the Way Towards More Efficient Systems  

    To identify 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 measured values of radiation counts. The Maximum Likelihood-Expectation Maximization (ML-EM) method and the Moore-Penrose Matrix Inverse (MPMI) method are examined. The ML-EM method has been used for image reconstruction of computed tomography, and the MPMI method is one of the solution methods for simultaneous linear equations. A simple calculation model simulating the containment vessel were constructed including detectors and radiation sources. In actual situation, 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) are smaller than positions of radiation source. 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 a radiation count at a detector and a radioactive source. 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 radiation count measurement positions. The ML-EM and the MPMI methods show different prediction accuracy depending on the calculation condition. It is found that the detectors should be placed at vicinity of radiation sources of interest and that the applicability of the inverse estimation does not strongly depend on the radioactivity distribution.

    Scopus

  95. Cache efficient flux region assignment for the method of characteristics

    Yamamoto A., Giho A., Endo T.

    International Conference on Physics of Reactors, PHYSOR 2018: Reactor Physics Paving the Way Towards More Efficient Systems   Part F168384-2 巻   頁: 788 - 799   2018年

     詳細を見る

    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 Physics of Reactors, PHYSOR 2018: Reactor Physics Paving the Way Towards More Efficient Systems  

    A flux region assignment algorithm to increase cache efficiency for the method of characteristics is proposed. In order to minimize the stride of memory access, flux region IDs are assigned based on the ray trace sequence during the method of characteristics calculation. The present method is implemented in the three-dimensional transport code GENESIS and its performance is confirmed through verification calculations ranging from single PWR fuel assembly to PWR full core benchmark problems. The present method can reduce computational time by improving cache efficiency while suppressing memory requirement.

    Scopus

  96. Development of reduced order model of severe accident analysis code for probabilistic safety margin analysis

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

    International Conference on Physics of Reactors, PHYSOR 2018: Reactor Physics Paving the Way Towards More Efficient Systems   Part F168384-5 巻   頁: 3042 - 3053   2018年

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    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 Physics of Reactors, PHYSOR 2018: Reactor Physics Paving the Way Towards More Efficient Systems  

    For probabilistic safety margin analysis, we developed a reduced order model (ROM) that reconstruct calculation results of typical severe accident progressions obtained by the severe accident analysis code, MAAP. The ROM is applied for fast reconstruction of time series plant parameters (e.g., pressure and temperature) obtained by the MAAP code. The ROM is applied to two severe accident scenarios, i.e., the station black out (SBO) with loss of all feed water capabilities and the large break loss of coolant accident (large break LOCA) without the emergency core cooling system (ECCS) capability. Verification results indicate that the ROM reasonably reproduces temporal variation of plant parameters with a few bases obtained by the ROM, which enables very fast reconstruction of complicated accident progression of a severe accident.

    Scopus

  97. Development of the uncertainty quantification method of activation in reactor structures using reduced-order modeling

    Yokoi K., Endo T., Yamamoto A., Hayashi K., Mizuno R., Kimura Y.

    International Conference on Physics of Reactors, PHYSOR 2018: Reactor Physics Paving the Way Towards More Efficient Systems   Part F168384-3 巻   頁: 1793 - 1804   2018年

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    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 Physics of Reactors, PHYSOR 2018: Reactor Physics Paving the Way Towards More Efficient Systems  

    Uncertainty quantification of activation in structure materials around a nuclear reactor is important for efficient planning of decommissioning. In order to quantify the uncertainty of activation in reactor structures, neutron shielding calculations are necessary. Since the neutron shielding calculation is performed for spatially large regions around a reactor, it requires a lot of computation time especially in the case of multi-dimensional geometry. In this study, the reduced-order modeling (ROM) is applied to the sensitivity and uncertainty analysis of activation in reactor structures to reduce computation time. The ROM can reduce calculation cost for evaluation of sensitivity coefficients by identifying important or sensitive subspace (active subspace, AS) in input data. In this study, an AS is constructed by the several sensitivity coefficients in reactor structures which are evaluated by the perturbation theory (PT). Sensitivity coefficients of activation throughout reactor structures are estimated using the AS and the uncertainty of activation is evaluated by the "sandwich formula." The calculation results indicate that the uncertainty of the activation in reactor structures can be reproduced with low calculation cost (approximately 30 neutron transport calculations) using the ROM in one-dimensional geometry of a 500 MWe class BWR.

    Scopus

  98. Estimation of region-wise even-parity discontinuity factor for MOC through iterative procedure

    Yamamoto A., Giho A., Endo T.

    International Conference on Physics of Reactors, PHYSOR 2018: Reactor Physics Paving the Way Towards More Efficient Systems   Part F168384-3 巻   頁: 1892 - 1903   2018年

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    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 Physics of Reactors, PHYSOR 2018: Reactor Physics Paving the Way Towards More Efficient Systems  

    To reduce angular and spatial discretization error of MOC with a coarse calculation condition, region-wise even-parity discontinuity factor (EPDF) for transport calculations is evaluated through an iterative procedure using only region-wise scalar flux, i.e., without odd parity angular flux, partial-, or net-current at region boundary. Region-wise EPDF is evaluated in a single assembly geometry with reflective boundary condition. The evaluated EPDFs are applied to a 2x2 colorset assembly configuration and its performance is compared to the conventional superhomogenization (SPH) method. The calculation results indicate that 1) no convergence issue is observed during the iteration process to estimate EPDF, 2) performance of the region-wise EPDF is better than that of the conventional SPH method.

    Scopus

  99. Estimation of subcriticality in dollar units based on integral method for arbitrary state-change in subcritical system

    Nonaka A., Endo T., Yamamoto A.

    Transactions of the American Nuclear Society   119 巻   頁: 1112 - 1115   2018年

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    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:Transactions of the American Nuclear Society  

    Scopus

  100. Estimation of subcriticality in dollar units using integral method for subcritical system

    Nonaka A., Endo T., Yamamoto A.

    International Conference on Physics of Reactors, PHYSOR 2018: Reactor Physics Paving the Way Towards More Efficient Systems   Part F168384-5 巻   頁: 3271 - 3282   2018年

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    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 Physics of Reactors, PHYSOR 2018: Reactor Physics Paving the Way Towards More Efficient Systems  

    A subcriticality measurement method using the integral method is developed for any stepwise transient. The integral method usually applied to a negative reactivity insertion using the rod-drop and the source-jerk methods. In this study, it is clarified that the integral method can be applied to measure positive reactivity insertion. During refueling of a reactor, not only reactivity but also neutron source intensity and neutron generation time simultaneously change. In order to accurately monitor the subcriticality under these conditions, variations of these parameters must be taken into consideration. Therefore, a subcriticality measurement method is developed, which is applicable to not only variation of reactivity but also simultaneous variations of neutron source intensity and neutron generation time. Furthermore, the present method utilizes only measurement value of neutron count rate, thus the proposed method is practical. Verification calculations are carried out for a step-wise change of reactivity, neutron source intensity, and neutron generation time. The results indicate that reactivity is accurately predicted for the change of the neutron source intensity, which is difficult to achieve by the neutron source multiplication method. On the other hand, although the change of the neutron generation time has considerable impacts on the estimated result of subcriticality, the predicted error is less than 10% for very large variation of neutron generation time.

    Scopus

  101. Estimation of subcriticality using particle filter method

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

    Transactions of the American Nuclear Society   118 巻   頁: 851 - 854   2018年

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    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:Transactions of the American Nuclear Society  

    Scopus

  102. Estimation of subcriticality using particle filter method

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

    AISTech - Iron and Steel Technology Conference Proceedings   2018-May 巻   頁: 851 - 854   2018年

     詳細を見る

    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:AISTech - Iron and Steel Technology Conference Proceedings  

    Scopus

  103. Estimation of Modeling Approximation Error of Core Analysis Using the Surrogate Model with Kriging 査読有り

    Tomomi Hanai, Tomohiro Endo, Yasuhiro Kodama, Yasunori Ohoka, Akio Yamamoto

    rans. Am. Nucl. Soc.   117 巻   頁: 1269-1272   2017年11月

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    記述言語:英語  

  104. Application of the GENESIS Code to the Kobayashi 3D Benchmark Problem 査読有り

    Akio Yamamoto, Akinori Giho, Tomohiro Endo

    Trans. Am. Nucl. Soc.   117 巻   頁: 1403-1406   2017年11月

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    担当区分:筆頭著者   記述言語:英語  

  105. Recent developments in the GENESIS code based on the Legendre polynomial expansion of angular flux method

    Yamamoto Akio, Giho Akinori, Endo Tomohiro

    NUCLEAR ENGINEERING AND TECHNOLOGY   49 巻 ( 6 ) 頁: 1143 - 1156   2017年9月

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    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:Nuclear Engineering and Technology  

    This paper describes recent development activities of the GENESIS code, which is a transport code for heterogeneous three-dimensional geometry, focusing on applications to reactor core analysis. For the treatment of anisotropic scattering, the concept of the simplified Pn method is introduced in order to reduce storage of flux moments. The accuracy of the present method is verified through a benchmark problem. Next, the iteration stability of the GENESIS code for the highly voided condition, which would appear in a severe accident (e.g., design extension) conditions, is discussed. The efficiencies of the coarse mesh finite difference and generalized coarse mesh rebalance acceleration methods are verified with various stabilization techniques. Use of the effective diffusion coefficient and the artificial grid diffusion coefficients are found to be effective to stabilize the acceleration calculation in highly voided conditions.

    DOI: 10.1016/j.net.2017.06.016

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  106. Uncertainty Quantification of Activation Due to Cross Section Data in Neutron Shielding Calculation 査読有り

    Kimihiro Yokoi,Tomohiro Endo,Akio Yamamoto,Ryoji Mizuno,Yoshio Kimura

    Proc. 2017 International. Cogress on Advances in Nuclear Power Plants     2017年4月

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    記述言語:英語  

    DOI: CD-ROM

  107. Application of Bias Factor Method using Random Sampling Technique for Critical Eigenvalue Prediction of BWR 査読有り

    Motohiro Ito,Tomohiro Endo,Akio Yamamoto,Yusuke Kuroda,Takashi Yoshii

    Proc. 2017 International Cogress on Advances in Nuclear Power Plants     2017年4月

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    記述言語:英語  

    DOI: CD-ROM

  108. Development of Dynamic Probabilistic Risk Assessment Model for PWR using Simplified Plant Simulation Method 査読有り

    Shohei Otsuki,Tomohiro Endo,Akio Yamamoto

    Proc. 2017 International Cogress on Advances in Nuclear Power Plants     2017年4月

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    記述言語:英語  

    DOI: CD-ROM

  109. Development of GENESIS, a Three-dimensional Heterogeneous Transport Code based on the LEAF Method 査読有り

    Akio Yamamoto,Akinori Giho,Tomohiro Endo

    Proc. Int. Conf. on Math. and Comp. Methods Applied to Nucl. Sci. & Eng. (M&C2017)     2017年4月

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    担当区分:筆頭著者   記述言語:英語  

    DOI: USB-DRIVE

  110. Theoretical Discussion of Statistical Error of Variance-to-Mean Ratio 査読有り

    Tomohiro Endo,Akio Yamamoto

    Proc. Int. Conf. on Math. and Comp. Methods Applied to Nucl. Sci. & Eng. (M&C2017)     2017年4月

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    記述言語:英語  

    DOI: USB-DRIVE

  111. Inverse Estimation of Unknown Radioactive Source using Detection Probability by Adjoint Calculation

    Shinji Sugaya,Tomohiro Endo,Akio Yamamoto

    Proc. Int. Conf. on Math. and Comp. Methods Applied to Nucl. Sci. & Eng. (M&C2017)     2017年4月

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    記述言語:英語  

    DOI: USB-DRIVE

  112. Uncertainty Quantification of Bell Factor for Sjöstrand Method due to Cross-section Data for MA core 査読有り

    Toshiki Kimura,Tomohiro Endo,Akio Yamamoto

    Proc. Int. Conf. on Math. and Comp. Methods Applied to Nucl. Sci. & Eng. (M&C2017)     2017年4月

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    記述言語:英語  

    DOI: USB-DRIVE

  113. A coupling model for the two-stage core calculation method with subchannel analysis for boiling water reactors

    Mitsuyasu Takeshi, Aoyama Motoo, Yamamoto Akio

    ANNALS OF NUCLEAR ENERGY   102 巻   頁: 77 - 84   2017年4月

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    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: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 core analysis model coupled with subchannel analysis within the two-stage calculation scheme using an assembly-based thermal-hydraulics calculation in the core analysis. The model changes the 2D lattice physics scheme, and couples with 3D subchannel analysis which evaluates the thermal-hydraulics characteristics within the coolant flow area divided as some subchannel regions. In order to couple with these two analyses, some BWR fuel assembly parameters are assumed and verified. The developed model is evaluated for the heterogeneous problem with and without a control rod. The present model is especially effective for the control rod inserted condition. The present model can incorporate the subchannel effect into the current two-stage core calculation method.

    DOI: 10.1016/j.anucene.2016.11.045

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  114. GENESIS: A Three-Dimensional Heterogeneous Transport Solver Based on the Legendre Polynomial Expansion of Angular Flux Method

    Yamamoto Akio, Giho Akinori, Kato Yuki, Endo Tomohiro

    NUCLEAR SCIENCE AND ENGINEERING   186 巻 ( 1 ) 頁: 1 - 22   2017年4月

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    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:Nuclear Science and Engineering  

    A heterogeneous transport solver in three-dimensional (3-D) geometry, GENESIS, is developed incorporating recent developments in the method of characteristics (MOC) in 3-D geometry. The Legendre Polynomial Expansion of Angular Flux (LEAF) method is implemented in the GENESIS code, in which neutron transport is calculated in two-dimensional (2-D) characteristics planes rather than in one-dimensional characteristics lines adopted in the conventional approach of 3-D MOC. Unlike the planar MOC method that combines 2-D MOC calculations through axial leakages, the GENESIS code explicitly considers angular and spatial dependence of outgoing and incoming angular fluxes between axial planes. Thus, the GENESIS code eliminates a crucial approximation used in the planarMOC method: No approximation is used for axial leakage treatment. The GENESIS code can handle flexible shapes of objects in rectangular or hexagonal geometry. A two-level, multigroup generalized coarse mesh rebalance acceleration method is adopted for efficient convergence of neutron transport calculation. Performance of the GENESIS code is verified through various benchmark calculations. The calculation results indicate the fidelity of the GENESIS code based on the LEAF method.

    DOI: 10.1080/00295639.2016.1273002

    Web of Science

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  115. Estimation of modeling approximation errors using data assimilation with the minimum variance approach

    Yamamoto Akio, Kinoshita Kuniharu, Endo Tomohiro

    JOURNAL OF NUCLEAR SCIENCE AND TECHNOLOGY   54 巻 ( 4 ) 頁: 459 - 471   2017年4月

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    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:Journal of Nuclear Science and Technology  

    In this paper, we estimate prediction errors owing to approximations in calculation models (modeling approximation error) using the data assimilation method. Correlations between the modeling approximation error and neutronics parameters obtained through calculations are evaluated in test configurations and then the evaluated correlations are used to predict the modeling approximation errors in design configuration. Formulae to estimate the modeling approximation error using the correlations are derived based on the minimum variance approach and the physical interpretation of the formulae is discussed through simple cases. The proposed method is applied in 2 × 2 and 3 × 3 fuel assembly geometries using specifications of the KAIST benchmark problem. The correlation between the modeling approximation error and parameters (neutron leakage in each fuel assembly) is estimated in 2 × 2 fuel assemblies and then the modeling approximation error in 3 × 3 fuel assemblies is predicted using the correlation. The calculation results not only indicate feasibility of the present method, but also suggest a need for further investigation on the assumptions used in the present study, i.e. applicability and robustness of the correlation among different geometries.

    DOI: 10.1080/00223131.2017.1286271

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  116. Automated generation of burnup chain for reactor analysis applications

    Tran V. -P., Tran H. -N., Yamamoto A., Endo T.

    KERNTECHNIK   82 巻 ( 2 ) 頁: 196 - 205   2017年4月

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    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:Kerntechnik  

    This paper presents the development of an automated generation of burnup chain for reactor analysis applications. Algorithms are proposed to reevaluate decay modes, branching ratios and effective fission product (FP) cumulative yields of a given list of important FPs taking into account intermediate reactions. A new burnup chain is generated using the updated data sources taken from the JENDL FP decay data file 2011 and Fission yields data file 2011. The new burnup chain is output according to the format for the SRAC code system. Verification has been performed to evaluate the accuracy of the new burnup chain. The results show that the new burnup chain reproduces well the results of a reference one with 193 fission products used in SRAC. Burnup calculations using the new burnup chain have also been performed based on UO2 and MOX fuel pin cells and compared with a reference chain th2cm6fp193bp6T.

    DOI: 10.3139/124.110671

    Web of Science

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  117. Estimation of Modeling Approximation Errors using Data Assimilation with the Minimum Variance Approach 査読有り

    A. Yamamoto,K. Kinoshita,T. Endo

    J. Nucl. Sci. Technol.   54 巻   頁: 459-471   2017年2月

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    担当区分:筆頭著者   記述言語:英語   掲載種別:研究論文(学術雑誌)  

  118. GENESIS - A Three-dimensional Heterogeneous Transport Solver based on the Legendre Polynomial Expansion of Angular Flux Method 査読有り

    A. Yamamoto,A. Giho,Y. Kato,T. Endo

    Nucl. Sci. Eng.   186 巻   頁: 1-22   2017年

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    担当区分:筆頭著者   記述言語:英語   掲載種別:研究論文(学術雑誌)  

  119. Application of the GENESIS Code to the Kobayashi 3D Benchmark Problem 査読有り

    Akio Yamamoto,Akinori Giho,Tomohiro Endo

    Trans. Am. Nucl. Soc.   117 巻   頁: 1403-1406   2017年

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    担当区分:筆頭著者   記述言語:英語  

  120. Estimation of Modeling Approximation Error of Core Analysis Using the Surrogate Model with Kriging 査読有り

    Tomomi Hanai,Tomohiro Endo,Yasuhiro Kodama,Yasunori Ohoka,Akio Yamamoto

    Trans. Am. Nucl. Soc.   117 巻   頁: 1269-1272   2017年

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    記述言語:英語  

  121. Application of the Bias Factor Method Using the Random Sampling Technique for Prediction Accuracy Improvement of Neutronics Parameters of BWR 査読有り

    Motohiro Ito,Tomohiro Endo,Akio Yamamoto,Yusuke Kuroda,Takashi Yoshii

    Trans. Am. Nucl. Soc.   117 巻   頁: 804-807   2017年

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    記述言語:英語  

  122. Recent developments in the GENESIS code based on the Legendre polynomial expansion of angular flux method 査読有り

    A. Yamamoto,A. Giho,T. Endo

    Nucl. Eng. Technol.   49 巻   頁: 1143-1156   2017年

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    担当区分:筆頭著者   記述言語:英語   掲載種別:研究論文(学術雑誌)  

  123. Automated Generation of Burnup Chain for Reactor Analysis Applications 査読有り

    V. P. Tran,H. N. Tran,A. Yamamoto,T. Endo

    Kerntechnik   82 巻   頁: 196-205   2017年

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    記述言語:英語   掲載種別:研究論文(学術雑誌)  

  124. A Coupling Model for the Two-stage Core Calculation Method with Subchannel Analysis for Boiling Water Reactors 査読有り

    T. Mitsuyasu, M. Aoyama, A. Yamamoto

    Ann. Nucl. Energy   102 巻   頁: 77-84   2017年

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    記述言語:英語   掲載種別:研究論文(学術雑誌)  

  125. Estimation of Sensitivity Coefficients of Core Characteristics based on Reduced-order Modeling using Sensitivity Matrix of Assembly Characteristics 査読有り

    R. Katano,T. Endo,A. Yamamoto,M. Abdo,H. Abdel-Khalik

    J. Nucl. Sci. Technol.   54 巻   頁: 637-647   2017年

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    記述言語:英語   掲載種別:研究論文(学術雑誌)  

  126. An efficient execution of Monte Carlo simulation based on delta-tracking method using GPUs

    Okubo Takuya, Endo Tomohiro, Yamamoto Akio

    JOURNAL OF NUCLEAR SCIENCE AND TECHNOLOGY   54 巻 ( 1 ) 頁: 30 - 38   2017年

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    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:Journal of Nuclear Science and Technology  

    An efficient execution method for Monte Carlo simulation using graphic processing unit (GPU) is proposed. The delta-tracking method is used since the delta-tracking method can reduce conditional branches and complexity of code implementation, which degrade computational performance on GPUs. In order to improve parallel efficiency in the eigenvalue calculation, generated fission neutrons are recorded using the atomic operation which avoids the data race in GPUs. We also propose a method to efficiently tally neutron flux in a region. The present method is based on the atomic operation and use of fixed-point type number instead of common floating-point type number. The verification calculations using the C5G7 benchmark problem show effectiveness of the proposed numerical algorithms on GPUs through comparison with calculations using central processing units.

    DOI: 10.1080/00223131.2016.1202793

    Web of Science

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  127. Estimation of sensitivity coefficients of core characteristics based on reduced-order modeling using sensitivity matrix of assembly characteristics

    Katano Ryota, Endo Tomohiro, Yamamoto Akio, Abdo Mohammad, Abdel-Khalik Hany

    JOURNAL OF NUCLEAR SCIENCE AND TECHNOLOGY   54 巻 ( 6 ) 頁: 637-647   2017年

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    記述言語:英語   掲載種別:研究論文(学術雑誌)  

    DOI: 10.1080/00223131.2017.1299052

    Web of Science

  128. Uncertainty quantification of activation due to cross section data in neutron shielding calculation

    Yoko K., Endo T., Yamamoto A., Mizuno R., Kimura Y.

    2017 International Congress on Advances in Nuclear Power Plants, ICAPP 2017 - A New Paradigm in Nuclear Power Safety, Proceedings     2017年

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    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:2017 International Congress on Advances in Nuclear Power Plants, ICAPP 2017 - A New Paradigm in Nuclear Power Safety, Proceedings  

    Quantification of activation in structure materials around a nuclear reactor using results of neutron shielding calculations is necessary for efficient planning of decommissioning. In this study, the random sampling method is applied to the neutron shielding calculation in order to estimate uncertainties of neutron flux and activation of 59Co in structure materials due to cross section covariance. The calculation results indicate that the magnitude of uncertainty of activation of 59Co due to cross section data is approximately 15∼30% inside structure materials.

    Scopus

  129. Evaluation of the n/γ discrimination performance of the neutron detector with eu doped TRUST-LiCaAlF6

    Maeno K., Endo T., Yamamoto A.

    2017 International Congress on Advances in Nuclear Power Plants, ICAPP 2017 - A New Paradigm in Nuclear Power Safety, Proceedings     2017年

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    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:2017 International Congress on Advances in Nuclear Power Plants, ICAPP 2017 - A New Paradigm in Nuclear Power Safety, Proceedings  

    The n/γ discrimination performance of the prototype neutron detector using Eu doped TRUST-LiCaAlF6 is quantitatively investigated, in order to confirm its applicability to the subcriticality monitoring in the decommissioning works at Fukushima Daiichi Nuclear Power Plant units 1-3, especially focusing on removal of fuel debris. In this paper, the experimental results of pulse-height distribution measurement using 252Cf at the Nagoya University Cobalt 60 irradiation facility are shown. Additionally, in order to investigate the applicability to the reactor noise analysis, preliminary experimental results of the Feynman-α method are also shown.

    Scopus

  130. Application of the genesis code to the Kobayashi 3D benchmark problem

    Yamamoto A., Giho A., Endo T.

    Transactions of the American Nuclear Society   117 巻   頁: 1403 - 1406   2017年

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    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:Transactions of the American Nuclear Society  

    Scopus

  131. Development of dynamic probabilistic risk assessment model for PWR using simplified plant simulation method

    Otsuki S., Endo T., Yamamoto A.

    2017 International Congress on Advances in Nuclear Power Plants, ICAPP 2017 - A New Paradigm in Nuclear Power Safety, Proceedings     2017年

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    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:2017 International Congress on Advances in Nuclear Power Plants, ICAPP 2017 - A New Paradigm in Nuclear Power Safety, Proceedings  

    The CMMC (Continuous Markov Monte Carlo) coupling method is proposed for quantifying accident scenarios that have time-dependence in accident progression and inter-dependence of individual events. In order to analyze large number of samples within realistic computing time, we developed a simplified accident progression analysis model and coupled the simplified model to the CMMC method. Large number of samples for station black out in PWR were analyzed using the random sampling method based on the CMMC method considering uncertainty of decay heat from a core. As a result, uncertainties of event initiation time and FPs release ratio to the atmosphere are estimated.

    Scopus

  132. Effective use of engineering reactor simulator for education of nuclear safety

    Yamamoto A., Endo T.

    2017 International Congress on Advances in Nuclear Power Plants, ICAPP 2017 - A New Paradigm in Nuclear Power Safety, Proceedings     2017年

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    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:2017 International Congress on Advances in Nuclear Power Plants, ICAPP 2017 - A New Paradigm in Nuclear Power Safety, Proceedings  

    Effective use of a reactor simulator to lecture nuclear safety for undergraduate/master's course students majoring nuclear engineering is discussed. Various scenarios from normal operating to severe accident conditions are used for an exercise of a few days' course. Through discussions on plant behaviors of various scenarios within a small group and among class members, participants can understand physics behind responses of a nuclear power plant.

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  133. Estimation of modeling approximation error of core analysis using the surrogate model with kriging

    Hanai T., Endo T., Kodama Y., Ohoka Y., Yamamoto A.

    Transactions of the American Nuclear Society   117 巻   頁: 1269 - 1272   2017年

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    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:Transactions of the American Nuclear Society  

    Scopus

  134. GENESIS - a Transport Solver in Three-Dimensional Heterogeneous Geometry based on the Leaf Method 査読有り

    A. Yamamoto,A. Giho,Y. Kato,T. Endo

    Proc. PHYSOR2016     2016年4月

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    担当区分:筆頭著者   記述言語:英語  

    DOI: DOI

  135. Statistical Error Estimation of the Feynman-α Method using the Bootstrap Method 査読有り

    T. Endo,A. Yamamoto,T. Yagi,C. H. Pyeon

    J. Nucl. Sci. Technol.   53 巻   頁: 1447-1453   2016年

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    記述言語:英語   掲載種別:研究論文(学術雑誌)  

  136. Discontinuity Factors for Simplified P3 Theory 査読有り

    A. Yamamoto,T. Sakamoto,T. Endo

    Nucl. Sci. Eng.   183 巻   頁: 39-51   2016年

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    担当区分:筆頭著者   記述言語:英語   掲載種別:研究論文(学術雑誌)  

  137. Bias Factor Method Using Random Sampling Technique 査読有り

    T. Endo,A. Yamamoto,T. Watanabe

    J. Nucl. Sci. Technol.   53 巻   頁: 1491-1501   2016年

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    記述言語:英語   掲載種別:研究論文(学術雑誌)  

  138. Cross Section Adjustment Methods based on Minimum Variance Unbiased Estimation 査読有り

    K. Yokoyama,A. Yamamoto

    J. Nucl. Sci. Technol.   53 巻   頁: 1622-1638   2016年

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    記述言語:英語   掲載種別:研究論文(学術雑誌)  

  139. A CMFD Acceleration Method for SP3 Advanced Nodal Method 査読有り

    A. Yamamoto,T. Sakamoto,T. Endo

    Nucl. Sci. Eng.   184 巻   頁: 168-173   2016年

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    担当区分:筆頭著者   記述言語:英語   掲載種別:研究論文(学術雑誌)  

  140. Reduction of MOC Discretization Errors Through a Minimization of Source Ratio Variances 査読有り

    M. Tabuchi,A. Yamamoto,E. Endo,M. Tatsumi

    J. Nucl. Sci. Technol.   53 巻   頁: 1858-1869   2016年

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    記述言語:英語   掲載種別:研究論文(学術雑誌)  

  141. An efficient execution of Monte Carlo simulation based on delta-tracking method using GPUs 査読有り

    T. Okubo,A. Yamamoto,E. Endo

    J. Nucl. Sci. Technol.   53 巻   頁: 1-9   2016年

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    記述言語:英語   掲載種別:研究論文(学術雑誌)  

  142. Uncertainty Quantification of Spatial Correction Factor for Sjöstrand Method due to Cross-Section Data 査読有り

    T. Kimura,T. Endo,A. Yamamoto

    Trans. Am. Nucl. Soc.   115 巻   頁: 1081-1084   2016年

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    記述言語:英語  

  143. Comparison of Fuel Loading Pattern Optimization Results using Exhaustive Search for Fresh Fuels and Local Search for Burned Fuels 査読有り

    S. Ishiguro,T. Endo,A. Yamamoto

    Trans. Am. Nucl. Soc.   115 巻   頁: 1265-1267   2016年

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    記述言語:英語  

  144. Comparison of the Numerical Stability between CMFD and GCMR with Stabilization Techniques 査読有り

    Akinori Giho,A. Yamamoto

    Trans. Am. Nucl. Soc.   115 巻   頁: 1245-1248   2016年

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    記述言語:英語  

  145. Application of Simplified Pn Approximation to Angular Distribution of Neutron Source in MOC Calculations

    A. Yamamoto,Akinori Giho,T. Endo

    Trans. Am. Nucl. Soc.   115 巻   頁: 1241-1244   2016年

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    担当区分:筆頭著者   記述言語:英語  

  146. Prediction on Underestimation of Statistical Uncertainty in Monte Carlo Eigenvalue Calculation for Two-Dimensional Heterogeneous Color Set Geometry 査読有り

    K. Hayashi,T. Endo,A. Yamamoto

    Trans. Am. Nucl. Soc.   115 巻   頁: 1213-1216   2016年

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    記述言語:英語  

  147. Uncertainty Quantification of Activation Due to Cross Section Data in Neutron Shielding Calculation 査読有り

    K. Yokoi,T. Endo,A. Yamamoto,R. Mizuno,Y. Kimura

    Trans. Am. Nucl. Soc.   115 巻   頁: 1085-1087   2016年

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    記述言語:英語  

  148. Development of a Simplified Estimation Method on Severe Accident Progression in PWR for Education 査読有り

    S. Otsuki,T. Endo,A. Yamamoto

    Trans. Am. Nucl. Soc.   113 巻   頁: 855-858   2015年11月

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    記述言語:英語  

  149. Application of Data Assimilation based on Bayesian Theory to Subcriticality Measurements using Area Ratio Method 査読有り

    K. Maeno,T. Endo,A. Yamamoto

    Trans. Am. Nucl. Soc.   113 巻   頁: 1282-1286   2015年11月

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    記述言語:英語  

  150. Theoretical Expression of Area Ratio Method Using Detected-Neutron Multiplication Factor 査読有り

    T. Endo,A. Yamamoto

    Trans. Am. Nucl. Soc.   113 巻   頁: 1208-1211   2015年11月

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    記述言語:英語  

  151. Application of Partially-Converged Solution of Assembly Calculation for Core Sensitivity Analysis based on Reduced Order Modeling 査読有り

    R. Katano,T. Endo,A. Yamamoto

    Trans. Am. Nucl. Soc.   113 巻   頁: 1161-1164   2015年11月

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    記述言語:英語  

  152. Underestimation of Statistical Uncertainty of Monte Carlo Method with Non-Analog of Fission Source Sampling 査読有り

    H. Hayashi,T. Endo,A. Yamamoto

    Trans. Am. Nucl. Soc.   113 巻   頁: 1153-1157   2015年11月

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    記述言語:英語  

  153. Uncertainty Estimation of Analysis Model using the Data Assimilation Method 査読有り

    K. Kinoshita,T. Endo,A. Yamamoto

    Trans. Am. Nucl. Soc.   113 巻   頁: 1141-1143   2015年11月

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    記述言語:英語  

  154. Discontinuity Factors for Simplified P3 Theory 査読有り

    A.Yamamoto,T. Sakamoto,T. Endo

    Proc. Reactor Physics Asia 2015 (RPHA15),Sep. 17-18, Jeju, Korea     2015年9月

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    担当区分:筆頭著者   記述言語:英語  

    DOI: CD-ROM

  155. Reactor Physics Activities in Nagoya University 査読有り

    A. Yamamoto,T. Endo

    Proc. Reactor Physics Asia 2015 (RPHA15), Sep. 17-18, Jeju, Korea     2015年9月

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    担当区分:筆頭著者   記述言語:英語  

    DOI: CD-ROM

  156. Development of New Statistical Geometry Model using the Delta-tracking Method 査読有り

    T. Koide,T. Endo,A. Yamamoto

    Proc. Reactor Physics Asia 2015 (RPHA15), Sep. 17-18, Jeju, Korea     2015年9月

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    記述言語:英語  

    DOI: CD-ROM

  157. Angular Dependent Transmission Probability Method for Fast Reactor Core Transport Analysis 査読有り

    A. Yamamoto,K. Kirimura,Y. Kamiyama,K. Yamaji,S. Kosaka,H. Matsumoto

    Trans. Am. Nucl. Soc.   112 巻   頁: 736-738   2015年6月

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    担当区分:筆頭著者   記述言語:英語  

  158. Development of MHI FBR Nuclear Design Code System CALAXY-H/ENSEMBLE-TRIZ 査読有り

    K. Kirimura,Y. Kamiyama,K. Yamaji,S. Kosaka,H. Matsumoto,A. Yamamoto

    Trans. Am. Nucl. Soc.   112 巻   頁: 733-735   2015年6月

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    記述言語:英語  

  159. Development of Core Sensitivity Analysis Based on Reduced-Order Modeling using Assembly Calculations 査読有り

    R. Katano,A. Yamamoto,T. Endo

    Trans. Am. Nucl. Soc.   112 巻   頁: 715-718   2015年6月

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    記述言語:英語  

  160. Efficient Execution of Monte Carlo Simulation Based on Pseudo-Scattering using GPU 査読有り

    T. Okubo,T. Endo,A. Yamamoto

    Trans. Am. Nucl. Soc.   112 巻   頁: 652-656   2015年6月

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    記述言語:英語  

  161. Application of correction technique using leakage index combined with SPH or discontinuity factors for energy collapsing on pin-by-pin BWR core analysis 査読有り

    T. Fujita,T. Endo,A. Yamamoto

    J. Nucl. Sci. Technol.   52 巻   頁: 355-370   2015年3月

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    記述言語:英語   掲載種別:研究論文(学術雑誌)  

  162. Uncertainty Quantification of LWR Core Characteristics using Random Sampling Method 査読有り

    A. Yamamoto,K. Kinoshita,T. Watanabe,E. Endo,Y. Kodama,T. Ohoka, T. Ushio, H. Nagano

    Nucl. Sci. Eng.   181 巻   頁: 160-174   2015年

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    担当区分:筆頭著者   記述言語:英語   掲載種別:研究論文(学術雑誌)  

  163. Integration of equivalence theory and ultra-fine-group slowing-down calculation for resonance self-shielding treatment in lattice physics code GALAXY 査読有り

    H. Koike,K. Yamaji,K. Kirimura,S. Kosaka,H. Matsumoto,A. Yamamoto

    J. Nucl. Sci. Technol.   52 巻   頁: 842-869   2015年

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    記述言語:英語   掲載種別:研究論文(学術雑誌)  

  164. Confidence interval estimation by bootstrap method for uncertainty quantification using random sampling method 査読有り

    T. Endo,T. Watanabe,A. Yamamoto

    J. Nucl. Sci. Technol.   52 巻   頁: 993-999   2015年

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    記述言語:英語   掲載種別:研究論文(学術雑誌)  

  165. Statistical Error Estimation Using Bootstrap Method for the Feynman-alpha Method 査読有り

    T. Endo,T. Shozawa,A. Yamamoto,C. H. Pyeon,T. Yagi

    Trans. Am. Nucl. Soc.   111 巻   頁: 1204-1207   2014年11月

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    記述言語:英語  

  166. Development of Overall Safety Evaluation Method for Operating Nuclear Power Plant Considering Aging Effects -Concept and Framework- 査読有り

    A. Yamamoto,T. Takata,K. Demachi,N. Sugiyama,A. Yamaguchi, H. Miyano

    Proc. ICMST-Kobe 2014     2014年11月

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    担当区分:筆頭著者   記述言語:英語  

    DOI: CD-ROM

  167. Development of Overall Safety Evaluation Method for Operating Nuclear Power Plant Considering Aging Effects - New Risk Indicators for Maintenance Procedure- 査読有り

    T. Takata,A. Yamaguchi,A. Yamamoto,K. Demachi,N. Sugiyama,H. Miyano

    Proc. ICMST-Kobe 2014     2014年11月

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    記述言語:英語  

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  168. Comparison of Spatial Source Expansion Methods in the Three Dimensional Transport Method LEAF 査読有り

    Y. Kato,T. Endo,A. Yamamoto

    Trans. Am. Nucl. Soc.   111 巻   頁: 1405-1408   2014年11月

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    記述言語:英語  

  169. Estimation of Sensitivity Coefficient using Random Sampling and L1-norm Minimization 査読有り

    T. Watanabe,T. Endo,A. Yamamoto,Y. Kodama,T. Ohoka,T. Ushio

    Trans. Am. Nucl. Soc.   111 巻   頁: 1391-1394   2014年11月

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    記述言語:英語   掲載種別:研究論文(学術雑誌)  

  170. Confidence Interval Estimation by Bootstrap Method for Uncertainty Quantification using Random Sampling 査読有り

    T. Endo,T. Watanabe,A. Yamamoto

    Proc. PHYSOR2014     2014年9月

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    記述言語:英語  

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  171. Impact of Nearest Neighbor Distribution of Fuel Particle on Neutronics Characteristics in Statistical Geometry Model 査読有り

    T. KoideT. Endo,A. Yamamoto,K. Kirimura,K. Yamaji

    Proc. PHYSOR2014     2014年9月

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    記述言語:英語  

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  172. Theoretical Prediction on Underestimation of Statistical Uncertainty for Fission Rate Tally in Monte Carlo Calculation 査読有り

    T. Endo.A. Yamamoto,K. Sakata

    Proc. PHYSOR2014     2014年9月

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    記述言語:英語  

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  173. Development of Legendre expansion of Angular Flux Method For 3D MOC Calculation 査読有り

    Y. Kato,A. Yamamoto,T. Endo

    Proc. PHYSOR2014     2014年9月

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    記述言語:英語  

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  174. Improvement of a Convergence Technique for MOC Calculation with Large Negative Self-Scattering Cross Section 査読有り

    M. Tabuchi,M. Tatsumi,A. Yamamoto,T. Endo

    Proc. PHYSOR2014     2014年9月

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    記述言語:英語  

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  175. Generation of Simplified Burnup Chain using Contribution Matrix of Nuclide Production 査読有り

    R. Katano,A. Yamamoto,T. Endo,Y. Kamiyama,K. Kirimura,S. Kosaka

    Proc. PHYSOR2014     2014年9月

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    記述言語:英語  

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  176. Uncertainty Quantification of Neutronics Characteristics using Latin Hypercube Sampling Method 査読有り

    K. Kinoshita,A. Yamamoto,T. Endo,Y. Kodama,Y. Ohoka,T. Ushio, H. Nagano

    Proc. PHYSOR2014     2014年9月

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    記述言語:英語  

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  177. Applicability of the Cross Section Adjustment Method based on Random Sampling Technique For Burnup 査読有り

    T. Watanabe,T. Endo,A. Yamamoto,Y. Kodama,Y. Ohoka,T. Ushio,

    Proc. PHYSOR2014     2014年9月

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    記述言語:英語  

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  178. Uncertainty Quantification of BWR Core Characteristics using Latin Hypercube Sampling Method 査読有り

    K. Kinoshita,A. Yamamoto,T. Endo,Y. Kodama,Y. Ohoka,T. Ushio,H. Nagano

    Proc. PHYSOR2014     2014年9月

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    記述言語:英語  

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  179. Investigation on Subcriticality Measurement using Inherent Neutron Source in Nuclear Fuel 査読有り

    T. Shiozawa,T. Endo,A. Yamamoto,C. H. Pyeon,T. Yagi

    Proc. PHYSOR2014     2014年9月

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    記述言語:英語  

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  180. Prediction on Underestimation of Variance for Fission Rate Distribution in Monte-Carlo Calculation 査読有り

    A. Yamamoto,K. Sakata,T. Endo

    Trans. Am. Nucl. Soc.   110 巻   頁: 515-518   2014年6月

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    担当区分:筆頭著者   記述言語:英語  

  181. An Improved CMFD Acceleration for SP3 Advanced Nodal Method 査読有り

    T. Sakamoto,A. Yamamoto,T. Endo

    Trans. Am. Nucl. Soc.   110 巻   頁: 535-537   2014年6月

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    記述言語:英語  

  182. Applicability of angular flux discontinuity factor preserving region-wise leakage for integro-differential transport equation 査読有り

    T. Sakamoto,T. Endo,A. Yamamoto

    J. Nucl. Sci. Technol.   51 巻   頁: 1264-1273   2014年6月

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    記述言語:英語   掲載種別:研究論文(学術雑誌)  

  183. Application of Augmented Reality to Nuclear Reactor Core Simulation for Fundamental Nuclear Engineering Education 査読有り

    K. Tsujita,T. Endo,A. Yamamoto

    Nucl. Technol.   185 巻   頁: 71-84   2014年

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    記述言語:英語   掲載種別:研究論文(学術雑誌)  

  184. Applicability of angular flux discontinuity factor preserving region-wise leakage for integro-differential transport equation 査読有り

    T. Sakamoto,T. Endo,A. Yamamoto

    J. Nucl. Sci. Technol.   51 巻   頁: 00-00   2014年

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    記述言語:英語   掲載種別:研究論文(学術雑誌)  

  185. A new technique for spectral interference correction on pin-by-pin BWR core analysis 査読有り

    T. Fujita,T. Endo,A. Yamamoto

    J. Nucl. Sci. Technol.   51 巻   頁: 783-797   2014年

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    記述言語:英語   掲載種別:研究論文(学術雑誌)  

  186. Cross Section Adjustment Method Based on Random Sampling Technique 査読有り

    T. Watanabe,T. Endo,A. Yamamoto,Y. Kodama,T. Ohoka,T. Ushio

    J. Nucl. Sci. Technol.   51 巻   頁: 590-599   2014年

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    記述言語:英語   掲載種別:研究論文(学術雑誌)  

  187. A macroscopic cross-section model for BWR pin-by-pin core analysis 査読有り

    T. Fujita,T. Endo,A. Yamamoto

    J. Nucl. Sci. Technol.   51 巻   頁: 282-304   2014年

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    記述言語:英語   掲載種別:研究論文(学術雑誌)  

  188. Subcriticality Measurement Technique using Inherent Neutron Source in Uranium Fuel 査読有り

    T. Shiozawa,T. Endo,A. Yamamoto,C. H. Pyeon,T. Yagi

    Trans. Am. Nucl. Soc.   109 巻   頁: 826-829   2013年11月

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    記述言語:英語  

  189. Estimation of Self-shielding Effect on Uncertainty of Neutronics Characteristics using Random Sampling Method and Continuous-energy Slowing-down Calculation 査読有り

    A. Yamamoto,S. Sato,T. Endo

    Trans. Am. Nucl. Soc.   109 巻   頁: 1436-1438   2013年11月

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    担当区分:筆頭著者   記述言語:英語  

  190. Uncertainty and Correlation Estimation of Reload Safety Parameters of PWR using Random Sampling Method 査読有り

    T. Watanabe,T. Endo,A. Yamamoto,Y. Ohoka,Y. Kodama,T. Ushio

    Trans. Am. Nucl. Soc.   109 巻   頁: 1365-1368   2013年11月

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    記述言語:英語  

  191. Behavior of Higher Order Fission Source Distribution in Monte-Carlo Calculations 査読有り

    A. Yamamoto,K. Sakata,T. Endo

    Trans. Am. Nucl. Soc.   109 巻   頁: 1361-1364   2013年11月

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    担当区分:筆頭著者   記述言語:英語  

  192. Evaluation of Higher Order Mode Components of Fission Source Distribution in Monte Carlo Calculation 査読有り

    K. Sakata,T. Endo,A. Yamamoto

    Proc. Int. Conf. SNA&MC2013, Paris, France, Oct. 27-31, 2013     2013年10月

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    記述言語:英語  

    DOI: DOI

  193. Higher Order Treatment on Temporal Derivative of Angular Flux for Time-Dependent MOC 査読有り

    K. Tsujita,T. Endo,A. Yamamoto,Y. Kamiyama,K. Kirimura

    Proc. Int. Conf. Math. and Comp. Methods Applied to Nucl. Sci. Eng. (M&C2013)     2013年5月

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    記述言語:英語  

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  194. Explicit Estimation of Higher Order Modes in Fission Source Distribution of Monte-Carlo Calculation 査読有り

    A. Yamamoto,K. Sakata,T. Endo

    Proc. Int. Conf. Math. and Comp. Methods Applied to Nucl. Sci. Eng. (M&C2013)     2013年5月

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    担当区分:筆頭著者   記述言語:英語  

    DOI: CD-ROM

  195. Reduction of Discretization Error for Ray Tracing of MOC Through a Correction on Collision Probabilities 査読有り

    M. Tabuchi,M. Tatsumi,A. Yamamoto,T. Endo

    Proc. Int. Conf. Math. and Comp. Methods Applied to Nucl. Sci. Eng. (M&C2013)     2013年5月

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    記述言語:英語  

    DOI: CD-ROM

  196. Application of the Multigrid Amplitude Function Method for Time-Dependent Transport Equation Using MOC 査読有り

    K. Tsujita,T. Endo,A. Yamamoto

    Proc. Int. Conf. Math. and Comp. Methods Applied to Nucl. Sci. Eng. (M&C2013)     2013年5月

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    記述言語:英語  

    DOI: CD-ROM

  197. Random Sampling-Based Cross-Section Adjustment Technique for LWR Core Analysis 査読有り

    S. Kato,T. Endo,A. Yamamoto

    Proc. ICAPP2013, Jeju, Korea Apr. 14-18, 2013     2013年4月

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    記述言語:英語  

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  198. Correction Technique for Course Group Cross Sections Considering Spectral Interference Effect on Pin-by-Pin BWR Core Analysis 査読有り

    T. Fujita,T. Endo,A. Yamamoto

    Proc. ICAPP2013, Jeju, Korea Apr. 14-18, 2013     2013年4月

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    記述言語:英語  

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  199. Convergence analysis of MOC inner iterations with large negative self-scattering cross-section 査読有り

    M. Tabuchi,A. Yamamoto,T. Endo,N. Sugimura

    J. Nucl. Sci. Technol.   50 巻   頁: 493-502   2013年

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    記述言語:英語   掲載種別:研究論文(学術雑誌)  

  200. Uncertainty estimation of core safety parameters using cross-correlations of covariance matrix 査読有り

    A. Yamamoto,Y. Yasue,T. Endo,M. Tatsumi,Y. Kodama,Y. Ohoka

    J. Nucl. Sci. Technol.   50 巻   頁: 966-978   2013年

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    担当区分:筆頭著者   記述言語:英語   掲載種別:研究論文(学術雑誌)  

  201. Preservation of transmission probabilities in the method of characteristics 査読有り

    M. Tabuchi,A. Yamamoto,T. Endo,N. Sugimura

    J. Nucl. Sci. Technol.   50 巻   頁: 837-843   2013年

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    記述言語:英語   掲載種別:研究論文(学術雑誌)  

  202. Study on Discontinuity Factor for Angular Flux in Transport Equation 査読有り

    T. Sakamoto,T. Endo,A. Yamamoto

    Trans. Am. Nucl. Soc.   108 巻   頁: 887-890   2013年

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    記述言語:英語  

  203. Estimation of Self-shielding Effect on Uncertainty of Neutronics Characteristics using Random Sampling Method and Continuous-energy Slowing-down Calculation 査読有り

    A. Yamamoto,S. Sato,T. Endo

    Trans. Am. Nucl. Soc.   109 巻   頁: 1436-1438   2013年

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    担当区分:筆頭著者   記述言語:英語  

  204. Uncertainty and Correlation Estimation of Reload Safety Parameters of PWR using Random Sampling Method 査読有り

    T. Watanabe,T. Endo,A. Yamamoto,Y. Ohoka,Y. Kodama,T. Ushio

    Trans. Am. Nucl. Soc.   109 巻   頁: 1365-1368   2013年

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    記述言語:英語  

  205. Behavior of Higher Order Fission Source Distribution in Monte-Carlo Calculations 査読有り

    A. Yamamoto,K. Sakata,T. Endo

    Trans. Am. Nucl. Soc.   109 巻   頁: 1361-1364   2013年

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    担当区分:筆頭著者   記述言語:英語  

  206. Subcriticality Measurement Technique using Inherent Neutron Source in Uranium Fuel 査読有り

    T. Shiozawa,T. Endo,A. Yamamoto,C. H. Pyeon,T. Yagi

    Trans. Am. Nucl. Soc.   109 巻   頁: 826-829   2013年

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    記述言語:英語  

  207. Few-Group Macroscopic Cross Section Adjustment for LWRs Using Random Sampling Technique 査読有り

    A. Yamamoto,S. Kato,T. Endo

    Trans. Am. Nucl. Soc.   108 巻   頁: 894-897   2013年

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    担当区分:筆頭著者   記述言語:英語  

  208. Utilization of Discontinuity Factor in Integro-differential Type of Boltzmann Transport Equation 査読有り

    A. Yamamoto

    Nucl. Sci. Eng   172 巻   頁: 259-267   2012年11月

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    担当区分:筆頭著者   記述言語:英語   掲載種別:研究論文(学術雑誌)  

  209. Development of Erbia Credit Super High Burnup Fuel: Evaluation of Minimum Erbia Content for Criticality Safety Analyses 査読有り

    M. Yamasaki,H. Unesaki,A. Yamamoto,T. Takeda,M. Mori

    Nucl. Technol.   180 巻   頁: 18-27   2012年10月

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    記述言語:英語   掲載種別:研究論文(学術雑誌)  

  210. Analysis of integral experiment on erbia-loaded thermal spectrum cores using Kyoto University Critical Assembly by MCNP code with various cross section libraries 査読有り

    Y. Tur,T. Endo,A. Yamamoto

    J. Nucl. Sci. Technol.   49 巻   頁: 1028-1041   2012年10月

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    記述言語:英語   掲載種別:研究論文(学術雑誌)  

  211. An Optimization Approach to Establish an Appropriate Energy Group Structure for BWR Pin-by-Pin Core Analysis 査読有り

    T. Fujita,K. Tada,T. Endo,A. Yamamoto,S. Kosaka,G. Hirano,K. Nozaki

    J. Nucl. Sci. Technol.   49 巻   頁: 689-707   2012年7月

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    記述言語:英語   掲載種別:研究論文(学術雑誌)  

  212. Advanced Resonance Self-Shielding Method for Gray Resonance Treatment in Lattice Physics Code GALAXY 査読有り

    H. Koike,K. Yamaji,K. Kirimura,D. Sato,H. Matsumoto,A. Yamamoto

    J. Nucl. Sci. Technol.   49 巻   頁: 725-747   2012年7月

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    記述言語:英語   掲載種別:研究論文(学術雑誌)  

  213. Efficient Fission Neutron Spectrum Matrix Representation by Singular Value Decomposition Technique 査読有り

    G. Chiba,A. Yamamoto,M. Tsuji,T. Narabayashi

    J. Nucl. Sci. Technol.   49 巻   頁: 748-753   2012年7月

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    記述言語:英語   掲載種別:研究論文(学術雑誌)  

  214. Analysis of Erbia-Loaded Critical Experiments in KUCA Using AEGIS Cross Section Library 査読有り

    A. Yamamoto,T. Endo,X. Wu

    Trans. Am. Nucl. Soc.   106 巻   頁: 715-718   2012年6月

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    担当区分:筆頭著者   記述言語:英語  

  215. Kinetic Calculation Method in Space-Time Frame Using Characteristic Line 査読有り

    K. Tsujita,T. Endo,A. Yamamoto

    Trans. Am. Nucl. Soc.   106 巻   頁: 743-746   2012年6月

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    記述言語:英語  

  216. A Unified Approach for Numerical Calculation of Space-dependent Kinetics Equation 査読有り

    Y. Ban,T. Endo,A. Yamamoto

    J. Nucl. Sci. Technol.   49 巻   頁: 496-515   2012年5月

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    記述言語:英語   掲載種別:研究論文(学術雑誌)  

  217. Uncertainty Estimation of Core Safety Parameters using Cross-Correlations of Covariance Matrix 査読有り

    A. Yamamoto,Y. Yasue,T. Endo,Y. Kodama,Y. Ohoka,M. Tatsumi

    Proc. Physor2012 - Advanced in Reactor Physics - Linking research, Industry and Education     2012年4月

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    担当区分:筆頭著者   記述言語:英語  

    DOI: DOI

  218. Estimation of average burnup of damaged fuels loaded in Fukushima Dai-ichi reactors by using the 134Cs/137Cs ratio method 査読有り

    T. Endo,S. Sato,A. Yamamoto

    Proc. Physor2012 - Advanced in Reactor Physics - Linking research, Industry and Education     2012年4月

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    記述言語:英語  

    DOI: DOI

  219. Multi-Physics Nuclear Reactor Simulator for Advanced Nuclear Engineering Education 査読有り

    A. Yamamoto

    Proc. Physor2012 - Advanced in Reactor Physics - Linking research, Industry and Education     2012年4月

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    担当区分:筆頭著者   記述言語:英語  

    DOI: DOI

  220. Development of Erbia Credit Super High Burnup Fuel: Experiments and Numerical Analyses 査読有り

    M. Yamasaki,H. Unesaki,A. Yamamoto,T. Takeda,M. Mori

    Nucl. Technol.   177 巻   頁: 63-72   2012年1月

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    記述言語:英語   掲載種別:研究論文(学術雑誌)  

  221. Subcriticality Measurement by Neutron Source Multiplication Method with Detected-Neutron Multiplication Factor 査読有り

    T. Endo,A. Yamamoto,C. H. Pyeon,T. Yagi

    Trans. Am. Nucl. Soc.,   107 巻   頁: 648-651   2012年

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    記述言語:英語  

  222. Correction of Spectral Interference Effect on Pin-by-Pin BWR Core Analysis 査読有り

    T. Fujita,T. Endo,A. Yamamoto

    Trans. Am. Nucl. Soc.   107 巻   頁: 1141-1143   2012年

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    記述言語:英語  

  223. Efficient Calculation Scheme with Preservation of Transmission Probabilities in the Method of Characteristics 査読有り

    M. Tabuchi,N. Sugimura,A. Yamamoto,T. Endo

    Trans. Am. Nucl. Soc.   107 巻   頁: 1105-1107   2012年

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    記述言語:英語  

  224. Higher order Treatment on Temporal Derivative of Angular Flux for Time-dependent MOC 査読有り

    K. Tsujita,E. Endo,A. Yamamoto,Y. Kamiyama,K. Kirimura

    Trans. Am. Nucl. Soc.   107 巻   頁: 1101-1104   2012年

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    記述言語:英語  

  225. Detected-Neutron Multiplication Factor Measured by Neutron Source Multiplication Method 査読有り

    T. Endo,A. Yamamoto,Y. Yamane

    Ann. Nucl. Energy   38 巻   頁: 2417-2427   2011年11月

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    記述言語:英語   掲載種別:研究論文(学術雑誌)  

  226. Evaluation of Correlation Among Uncertainties of Core Neutronic Parameters in LWR 査読有り

    Y. Yasue,T. Endo,A. Yamamoto,Y. Kodama,Y. Ohoka,M. Tatsumi

    Trans. Am. Nucl. Soc.   105 巻   頁: 486-488   2011年11月

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    記述言語:英語  

  227. Assembly Discontinuity Factor for Angular Flux in Transport Calculation 査読有り

    A. Yamamoto,T. Endo

    Trans. Am. Nucl. Soc.   105 巻   頁: 862-864   2011年11月

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    担当区分:筆頭著者   記述言語:英語  

  228. Development of a Lattice Physics Code for Sensitivity Analysis Based on Generalized Perturbation Theory 査読有り

    S. Kato,T. Endo,A. Yamamoto,Y. Kimura

    Trans. Am. Nucl. Soc.   105 巻   頁: 851-854   2011年11月

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    記述言語:英語  

  229. Application of the Discrete Ordinate CIP Scheme and the Fictitious Source Method for Reactor Shielding 査読有り

    K. Nakano, T. Endo, A. Yamamoto

    Proc. 19th International Conference On Nuclear Engineering (ICONE19)     頁: 0   2011年10月

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    記述言語:英語  

  230. A Systematic Approach to an Establish Energy Group Structure for BWR Pin-by-Pin Core Analysis 査読有り

    T. Fujita,K. Tada,T. Endo,A. Yamamoto,S. Kosaka,G. Hirano,K. Nozaki

    Proc. 19th International Conference On Nuclear Engineering (ICONE19)     頁: 0   2011年10月

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    記述言語:英語  

  231. Treatment of History Effect on Macroscopic Cross Section Model for BWR Pin-by-Pin Core Analysis 査読有り

    T. Fujita,K. Tada,T. Endo,A. Yamamoto,S. Kosaka,G. Hirano,K. Nozaki

    Proc. 19th International Conference On Nuclear Engineering (ICONE19)     頁: 0   2011年10月

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    記述言語:英語  

  232. A new calculation method for the generalized adjoint flux using the method of characteristics 査読有り

    S. Kato, T. Endo, A. Yamamoto

    Proc. 19th International Conference On Nuclear Engineering (ICONE19)     頁: 0   2011年10月

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    記述言語:英語  

  233. Application of Augmented Reality for Reactor Core Simulation 査読有り

    K. Tsujita, T. Endo, A. Yamamoto

    Proc. 19th International Conference On Nuclear Engineering (ICONE19)     頁: 0   2011年10月

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    記述言語:英語  

  234. Optimum In-Core Power Sharing with Multicycle Coupling Effect 査読有り

    A. Yamamoto,T. Iwata,Y. Yamane

    Prog. Nucl. Energy   53 巻   頁: 593-599   2011年8月

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    担当区分:筆頭著者   記述言語:英語   掲載種別:研究論文(学術雑誌)  

  235. Application of the Robust Design Concept for Fuel Loading Pattern," 査読有り

    T. Endo,K. Ohori,A. Yamamoto

    J. Nucl. Sci. Technol.   48 巻   頁: 1077-1086   2011年7月

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    記述言語:英語   掲載種別:研究論文(学術雑誌)  

  236. Improved Derivation of Multigroup Effective Cross Section for Heterogeneous System by Equivalence Theory 査読有り

    A. Yamamoto,T. Endo,G. Chiba

    Nucl. Sci. Eng   168 巻   頁: 75-92   2011年6月

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    担当区分:筆頭著者   記述言語:英語   掲載種別:研究論文(学術雑誌)  

  237. A Unified Numerical Algorithm for Space-Dependent Kinetic Equation 査読有り

    T. Endo,Y. Ban,A. Yamamoto

    Trans. Am. Nucl. Soc.   104 巻   頁: 865-867   2011年6月

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    記述言語:英語  

  238. A Derivation of Discontinuity Factor for Angular Flux in Integro-Differential Transport Equation 査読有り

    A. Yamamoto,T. Endo,Y. A. Chao

    Trans. Am. Nucl. Soc.   104 巻   頁: 815-817   2011年6月

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    担当区分:筆頭著者   記述言語:英語  

  239. Convergence Analysis of MOC with Large Negative Self-Scattering Cross Section 査読有り

    M. Tabuchi, H. Tagawa, A. Yamamoto, M. Tatsumi

    Trans. Am. Nucl. Soc.   104 巻   頁: 809-811   2011年6月

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    記述言語:英語  

  240. Resonance Calculation for Large and Complicated Geometry Using Tone's Method by Incorporating the Method of Characteristics 査読有り

    H. Yu,T. Endo,A. Yamamoto

    J. Nucl. Sci. Technol   48 巻   頁: 330-336   2011年3月

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    記述言語:英語   掲載種別:研究論文(学術雑誌)  

  241. Overview of Core Simulation Methodologies for Light Water Reactor Analysis 査読有り

    A. Yamamoto,T. Endo

    Int. J. Nucl. Safety and Simulation     頁: 12-21   2011年3月

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    担当区分:筆頭著者   記述言語:英語   掲載種別:研究論文(学術雑誌)  

  242. Explicit Time Integration Scheme using Krylov Subspace Method for Reactor Kinetics Equation 査読有り

    Y. Ban,T. Endo,A. Yamamoto,Y. Yamane

    J. Nucl. Sci. Technol.   48 巻   頁: 243-255   2011年2月

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    記述言語:英語   掲載種別:研究論文(学術雑誌)  

  243. Improvement of Tone's Method with Two-term Rational Approximation 査読有り

    A. Yamamoto,T. Endo,G. Chiba

    J. Nucl. Sci. Technol.   48 巻   頁: 263-271   2011年2月

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    担当区分:筆頭著者   記述言語:英語   掲載種別:研究論文(学術雑誌)  

  244. Application of Quick Subchannel Analysis Method for Three-Dimensional Pin-by-Pin BWR Core Calculations 査読有り

    K. Tada,T. Fujita,T. Endo,A. Yamamoto,S. Kosaka,G. Hirano,K. Nozaki

    J. Nucl. Sci. Technol   48 巻   頁: 1437-1452   2011年

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    記述言語:英語   掲載種別:研究論文(学術雑誌)  

  245. AEGIS: An Advanced Lattice Physics Code for Light Water Reactor Analyses 査読有り

    A. Yamamoto,T. Endo,M. Tabuchi,N. Sugimura,T. Ushio,M. Mori,M. Tatsumi,Y. Ohoka

    Nuclear Engineering and Technology   42 巻 ( 5 ) 頁: 500-519   2010年10月

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    担当区分:筆頭著者   記述言語:英語   掲載種別:研究論文(学術雑誌)  

  246. Incorporation of two-term rational approximation in Tone method for resonance calculation 査読有り

    A. Yamamoto,T. Endo,G. Chiba

    Trans. Am. Nucl. Soc.   103 巻   頁: 711-713   2010年6月

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    担当区分:筆頭著者   記述言語:英語  

  247. Effect of uncertainty of planned cycle length in multi-cycle fuel optimization 査読有り

    K. Ohori,T. Endo,A. Yamamoto

    Trans. Am. Nucl. Soc.   103 巻   頁: 707-710   2010年6月

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    記述言語:英語  

  248. Investigation of theoretical approach to establish energy group structure for BWR pin-by-pin core analysis 査読有り

    T. Fujita,K. Otsuka,K. Tada,T. Endo,A. Yamamoto,S. Kosaka,G. Hirano

    Trans. Am. Nucl. Soc.   103 巻   頁: 721-723   2010年6月

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    記述言語:英語  

  249. The Study on Erbia Credit Super-High-Burnup Fuel with Isotopically Modified Erbia 査読有り

    M. Yamasaki,H. Unesaki,A. Yamamoto

    Trans. Am. Nucl. Soc.   103 巻   頁: 735-736   2010年6月

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    記述言語:英語  

  250. Utilization of Discontinuity Factor in Integro-differential Type of Boltzmann Transport Equation 査読有り

    A. Yamamoto

    Proc. PHYSOR 2010 - Advances in Reactor Physics to Power the Nuclear Renaissance, Pittsburg, PA, May 2010   1 巻   2010年5月

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    担当区分:筆頭著者   記述言語:英語  

    DOI: DOI

  251. Resonance Calculation for Large and Complicated Geometry using Tone's Method by Incorporating Method of Characteristics 査読有り

    H. Yu,A. Yamamoto,Y. Yamane

    Proceedings of the 18th International Conference on Nuclear Engineering, ICONE18, Xian, China, May 2010   1 巻   2010年5月

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    記述言語:英語  

    DOI: CD-ROM

  252. A New Robust Cross Section Representation Methodology for PWR Core Simulator 査読有り

    D. Sato,S. Tsubota,K. Yamaji,H. Koike,H. Matsumoto,A. Yamamoto

    Proc. PHYSOR 2010 - Advances in Reactor Physics to Power the Nuclear Renaissance, Pittsburg, PA, May 2010   1 巻   2010年5月

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    記述言語:英語  

    DOI: CD-ROM

  253. A Resonance Calculation Method based on the Multi-Terms Rational Approximation for General Geometry with Gray Resonance Absorbers 査読有り

    H. Koike,K. Yamaji,D. Sato,S. Tsubota,H. Matsumoto,A. Yamamoto

    Proc. PHYSOR 2010 - Advances in Reactor Physics to Power the Nuclear Renaissance, Pittsburg, PA, May 2010   1 巻   2010年5月

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    記述言語:英語  

    DOI: CD-ROM

  254. Numerical Solution of Reactor Kinetics Equation with Krylov Subspace Method for Matrix Exponential 査読有り

    Y. Ban,A. Yamamoto,Y. Yamane

    Proc. PHYSOR 2010 - Advances in Reactor Physics to Power the Nuclear Renaissance, Pittsburg, PA, May 2010   1 巻   2010年5月

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    記述言語:英語  

    DOI: CD-ROM

  255. Investigation on Macroscopic Cross section Model for BWR Pin-by-pin Core Analysis 査読有り

    T. Fujita,K. Tada,A. Yamamoto,Y. Yamane,S. Kosaka,G. Hirano

    Proc. PHYSOR 2010 - Advances in Reactor Physics to Power the Nuclear Renaissance, Pittsburg, PA, May 2010   1 巻   2010年5月

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    記述言語:英語  

    DOI: CD-ROM

  256. Validation of Neutron Current Formulations for the Response Matrix Method based on the SP3 Theory 査読有り

    K. Tada,A. Yamamoto,Y. Yamane,S. Kosaka,G. Hirano

    Ann. Nucl. Energy   37 巻   頁: 22-27   2010年1月

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    記述言語:英語   掲載種別:研究論文(学術雑誌)  

  257. Measuring the Photoneutrons Originating from D(γ, n)H reaction after the Shutdown of an Operational BWR

    M Watanabe,A. Yamamoto,Y. Yamane

    J. Nucl. Sci. Technol   46 巻   頁: 1099-1112   2009年12月

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    記述言語:英語   掲載種別:研究論文(学術雑誌)  

  258. An Improved Inverse Analysis Model for Fuel Loading Pattern Optimization 査読有り

    H. N. Tran,A. Yamamoto,Y. Yamane

      46 巻   頁: 1162-1169   2009年12月

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    記述言語:英語   掲載種別:研究論文(学術雑誌)  

  259. *A New Framework of Resonance Calculation Method Based on the Sub-group Method (1); Theory

    A. Yamamoto,H. Koike,Y. Yamane

    Trans. Am. Nucl. Soc   100 巻   頁: 647-649   2009年11月

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    記述言語:英語  

  260. Examination of Pin-by-pin Fission Rate Distribution in Large Geometry Evaluated by the Monte-Carlo Method 査読有り

    A. Yamamoto,R. Nakamura

    Ann. Nucl. Energy   36 巻   頁: 1726-1733   2009年11月

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    担当区分:筆頭著者   記述言語:英語   掲載種別:研究論文(学術雑誌)  

  261. The Resonance Treatment in the AEGIS code

    N. Sugimura,M. Tabuchi,A. Yamamoto,M. Tatsumi

    Trans. Am. Nucl. Soc   100 巻   頁: 654-655   2009年11月

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    記述言語:英語  

  262. A New Framework of Resonance Calculation Method Based on the Sub-group Method (2); Calculation

    A. Yamamoto,H. Koike,Y. Yamane

    Trans. Am. Nucl. Soc   100 巻   頁: 650-651   2009年11月

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    記述言語:英語  

  263. Projected Predictor-corrector Method for Lattice Physics Burnup Calculations 査読有り

    A. Yamamoto,M. Tatsumi,N. Sugimura

    Nucl. Sci. Eng   163 巻   頁: 144-151   2009年10月

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    担当区分:筆頭著者   記述言語:英語   掲載種別:研究論文(学術雑誌)  

  264. Application of Discontinuity Factor for Integro-Differential Transport Equation

    A. Yamamoto

    Trans. Am. Nucl. Soc.   101 巻   頁: 402-404   2009年6月

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    記述言語:英語  

  265. Measurement of Reactivity Worth of Rare-Earth Elements at Kyoto University Critical Assembly

    H. Okochi,A. Yamamoto,Y. Yamane,T. Kitada,H. Unesaki

    Trans. Am. Nucl. Soc.   101 巻   頁: 737-738   2009年6月

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    記述言語:英語  

  266. Comparison of Partial Current Formulation for Response Matrix Method Based on SP3 Theory

    K. Tada,A. Yamamoto,Y. Yamane,S. Kosaka,G. Hirano

    Trans. Am. Nucl. Soc.   101 巻   頁: 717-719   2009年6月

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    記述言語:英語  

  267. An Improved Inverse Analysis Model for Core Calculation of Fuel Loading Pattern Optimization in LWRs

    H. N. Tran,A. Yamamoto,Y. Yamane

    Trans. Am. Nucl. Soc   101 巻   頁: 730-732   2009年6月

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    記述言語:英語  

  268. Subcriticality Estimation of Large FBR by Detectable Multiplication Factor kdet 査読有り

    K. Sugawara,Y. Yamane,A. Yamamoto,S. Okajima

    Trans. Am. Nucl. Soc.   101 巻   頁: 743-745   2009年6月

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    記述言語:英語  

  269. Verification of the Resonance Calculation Model for Rod Cluster Control Based on Ultra-fine-group Spectrum Calculation in the AEGIS code

    M. Tabuchi,N. Sugimura,T. Ushio,M. Mori,A. Yamamoto,M. Tatsumi,Y. Ohoka

    Proc. Int. Conf. Mathematics, Computational Methods & Reactor Physics (M&C2009)   CD-ROM 巻   2009年5月

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    記述言語:英語  

    DOI: CD-ROM

  270. Progress of Criticality Experiments and Nuclear Design Studies on Erbia-Bearing Super High Burnup Fuel 査読有り

    M. Yamasaki,T. Kuroishi,T. Takeda,A. Yamamoto,H. Unesaki,M. Mori,S. Sano

    Proc. International Congress on Advances in Nuclear Power Plants (ICAPP09),   CD-ROM 巻   2009年5月

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    記述言語:英語  

    DOI: CD-ROM

  271. Application of the SP3 nodal method with second order source and leakage approximations in axial direction for BWR pin-by-pin core analysis

    K. Tada,A. Yamamoto,Y. Yamane,S. Kosaka,G. Hirano

    Proc. Int. Congress on Advances in Nuclear Power Plants (ICAPP09)   CD-ROM 巻   2009年5月

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    記述言語:英語  

    DOI: CD-ROM

  272. Development of Multi-Stage Stochastic PWR Loading Pattern Search Code SAMPLS based on Search Space Division Method using Hamming Distance and Built-in Fresh Fuel Templates

    K. Ishitani,M. Adachi,J. Ueno,A. Yamamoto

    Proc. Advances in Nuclear Fuel Management IV (ANFM 2009)   CD-ROM 巻   2009年5月

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    記述言語:英語  

    DOI: CD-ROM

  273. Optimum In-Core Power Sharing with Multi-cycle Coupling Effect

    A. Yamamoto,T. Iwata,Y. Yamane

    Proc. Advances in Nuclear Fuel Management IV (ANFM 2009)   CD-ROM 巻   2009年4月

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    記述言語:英語  

    DOI: CD-ROM

  274. The outline of Development Project on Erbia Bearing Super-High Burnup Fuel

    M. Yamasaki,T. Kuroishi,T. Takeda,A. Yamamoto,H. Unesaki,T. Sano,M. Mori

    Proc. Advances in Nuclear Fuel Management IV (ANFM 2009)   CD-ROM 巻   2009年4月

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    記述言語:英語  

    DOI: CD-ROM

  275. Applicability of the Enhanced Neutron Current Method for Non-convex Fuel Shapes

    A. Yamamoto

    Ann. Nucl. Energy   36 巻   頁: 193-198   2009年3月

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    記述言語:英語   掲載種別:研究論文(学術雑誌)  

  276. Treatment of Staggered Mesh for BWR Pin-by-Pin Core Analysis

    K. Tada,A. Yamamoto,Y. Yamane

    J. Nucl. Sci. Technol   46 巻   頁: 163-174   2009年2月

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    記述言語:英語   掲載種別:研究論文(学術雑誌)  

  277. *Evaluation of Background Cross-section for Heterogeneous and Complicated Geometry by the Enhanced Neutron Current Method 査読有り

    A. Yamamoto

    J. Nucl. Sci. Technol   45 巻   頁: 1287-1292   2008年12月

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    担当区分:筆頭著者   記述言語:英語   掲載種別:研究論文(学術雑誌)  

  278. Applicability of the Diffusion and the Simplified P3 Theories for Pin-by-pin Geometry of BWR 査読有り

    K. Tada,A. Yamamoto,Y. Yamane,Y. Kitamura

    J. Nucl. Sci. Technol   45 巻   頁: 997-1008   2008年10月

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    記述言語:英語   掲載種別:研究論文(学術雑誌)  

  279. Development of AEGIS, a Next Generation Lattice Physics Code

    M. Tabuchi,N. Sugimura,A. Yamamoto,M. Tatsumi

    Proc. 16th Pacific Basin Nuclear Conference (16PBNC)   CD-ROM 巻   2008年10月

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    記述言語:英語  

    DOI: DOI

  280. Some properties of zero power neutron noise in a time-varying medium with delayed neutrons 査読有り

    Y. Kitamura,L. Pal,I. Pazsit,A. Yamamoto,Y. Yamane

    Ann. Nucl. Energy   35 巻   頁: 1621-1627   2008年9月

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    記述言語:英語   掲載種別:研究論文(学術雑誌)  

  281. Development of a Prototype Pin-by-pin Fine Mesh Calculation Code for BWR Core Analysis 査読有り

    K. Tada,A. Yamamoto,S. Kosaka,G. Hirano,Y. Yamane

    Proc. International Conference on the Physics of Reactors (Physor2008)     頁: CD-ROM   2008年9月

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    記述言語:英語  

  282. Verification of the AEGIS/SCOPE2 In-core Fuel Management System 査読有り

    M. Tatsumi,Y. Ohoka,N. Sugimura,A.Yamamoto

    Proc. International Conference on the Physics of Reactors (Physor2008)     頁: CD-ROM   2008年9月

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    記述言語:英語  

  283. Projected Predictor-corrector Method for Burnup Calculations of Gd-Bearing Fuel Assemblies 査読有り

    A. YamamotoM. Tatsumi,N. Sugimura

    Proc. International Conference on the Physics of Reactors (Physor2008)     頁: CD-ROM   2008年9月

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    担当区分:筆頭著者   記述言語:英語  

  284. Measuring the Ratio of 242Cm to 244Cm in Operating BWR Cores Using Start-Up Range Neutron Monitors 査読有り

    M. Watanabe,A. Yamamoto,Y. Yamane

    J. Nucl. Sci. Technol   45 巻   頁: 498-509   2008年6月

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    記述言語:英語   掲載種別:研究論文(学術雑誌)  

  285. Fast computation of the Neutron Transport Calculation with a Graphic Processor Unit (GPU) 査読有り

    Y. Kodama,A. Yamamoto,Y.Yamane, Y.Ohoka, M.Tatsumi

    Trans. Am. Nucl. Soc   99 巻   頁: 695-697   2008年6月

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    記述言語:英語  

  286. Optimization of Batch Power Sharing to Improve Discharge Burnup for Multicycle 査読有り

    T. Iwata,A. Yamamoto,Y.Yamane

    Trans. Am. Nucl. Soc   99 巻   頁: 703-705   2008年6月

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    記述言語:英語  

  287. Development of a Resonance Calculation Method Based on Discrete Treatment of Energy Ranges 査読有り

    H. Koike,A. Yamamoto,Y.Yamane

    Trans. Am. Nucl. Soc   99 巻   頁: 674-676   2008年6月

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    記述言語:英語  

  288. Evaluation of the Background Cross Section for Heterogeneous and Complicated Geometry by the Enhanced Neutron Current Method 査読有り

    A. Yamamoto,

    Trans. Am. Nucl. Soc   99 巻   頁: 671-673   2008年6月

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    担当区分:筆頭著者   記述言語:英語  

  289. Application of a Game Console for Fast Reactor Physics Computation 査読有り

    Y. Kodama,A. Yamamoto,Y. Yamane

    Trans. Am. Nucl. Soc   99 巻   頁: 698-699   2008年6月

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    記述言語:英語  

  290. Reduction of Spatial Discretization Error for the Method of Characteristics using the Mobile-Chord Ray Tracing Method 査読有り

    A. Yamamoto

    Ann. Nucl. Energy   35 巻   頁: 783-789   2008年5月

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  291. Development of Deterministic Code Based on the Discrete Ordinates Method for the Third-Order Neutron Correlation Technique 査読有り

    T. Endo,A. Yamamoto,Y. Yamane

    Ann. Nucl. Energy   35 巻   頁: 927-936   2008年5月

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    記述言語:英語   掲載種別:研究論文(学術雑誌)  

  292. Simplified Treatments of Anisotropic Scattering in LWR Core Calculations 査読有り

    A. YamamotoY. Kitamura,Y. Yamane

    J. Nucl. Sci. Technol   45 巻   頁: 217-229   2008年3月

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  293. Approximate Treatments of Anisotropic Scattering in LWR Analysis 査読有り

    A. Yamamoto

    Trans. Am. Nucl. Soc.   96 巻   頁: 505-507   2007年11月

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  294. Verification of Real-time Subcriticality Measurement based on Rossi-alpha Method Using Detection-time Acquisition System 査読有り

    S. Tsubota, Y. Yamane A. Yamamoto, Y. Kitamura

    Trans. Am. Nucl. Soc.   96 巻   頁: 625-626   2007年11月

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    記述言語:英語  

  295. Treatment of Staggered Mesh in BWR Pin-by-pin Fine Mesh Core Analysis 査読有り

    K. Tada, A. Yamamoto, M. Watanabe, H. Noda, Y. Kitamura, Y. Yamane

    Trans. Am. Nucl. Soc.   96 巻   頁: 508-510   2007年11月

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    記述言語:英語  

  296. A Barrier on the Public Communication of Nuclear Technology - How to Interpret Reactor Kinetics 査読有り

    A. Yamamoto

    Proc. International Symposium on EcoTopia Science 2007     頁: CD-ROM   2007年11月

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  297. Development of New Solid Angle Quadrature Sets to Satisfy Even- and Odd-Moment Conditions 査読有り

    T. Endo, A. Yamamoto

    J. Nucl. Sci. Technol.   44 巻   頁: 1249-1258   2007年10月

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    記述言語:英語   掲載種別:研究論文(学術雑誌)  

  298. Resonance Treatment Based on Ultra-fine-group Spectrum Calculation in the AEGIS Code 査読有り

    N. Sugimura, A. Yamamoto

    J. Nucl. Sci. Technol.   44 巻   頁: 958-966   2007年7月

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    記述言語:英語   掲載種別:研究論文(学術雑誌)  

  299. Accuracy of a Rapid Cell-Heterogeneous Calculation Method for LWR Core Analysis 査読有り

    A. Yamamoto

    Trans. Am. Nucl. Soc.   97 巻   頁: 582-584   2007年6月

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  300. Development of Er-SHB fuel: Critical Experiments and Analyses of Homogeneously Erbia-Loaded Cores in KUCA 査読有り

    Y. Goto, A. Yamamoto, H. Unesaki, T. Takeda, M. Mori, M. Yamasaki

    Trans. Am. Nucl. Soc.   97 巻   頁: 715-717   2007年6月

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    記述言語:英語  

  301. Applicability of the SP3 Nodal Method for BWR Pin-by-pin Core Analysis with Staggered Mesh 査読有り

    A. Yamamoto, K. Tada, Y. Kitamura, Y. Yamane

    Trans. Am. Nucl. Soc.   97 巻   頁: 569-572   2007年6月

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    担当区分:筆頭著者   記述言語:英語  

  302. AEGIS/SCOPE2, a Next-Generation In-core Fuel Management System (2) Core Calculation Code, SCOPE2 査読有り

    M. Tatsumi,H. Hyoudou,N. Sugimura,A. Yamamoto

    Trans. Am. Nucl. Soc   97 巻   頁: 562-564   2007年6月

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    記述言語:英語  

  303. Development of a New Lattice Physics Code GALAXY for Flexible Geometry Representation in NextGeneration Core Analyses System 査読有り

    K. Yamaji,H. Matsumoto,K. Kirimura,T. Takeda,A. Yamamoto

    Trans. Am. Nucl. Soc   97 巻   頁: 573-574   2007年6月

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    記述言語:英語  

  304. AEGIS/SCOPE2, a Next-Generation In-core Fuel Management System (1) Lattice Physics Code, AEGIS 査読有り

    N. Sugimura, T. Ushio, A. Yamamoto, M. Tatsumi

    Trans. Am. Nucl. Soc   97 巻   頁: 559-561   2007年6月

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    記述言語:英語  

  305. Calculation of higher moments of the neutron multiplication process in a time-varying medium 査読有り

    Y. Kitamura, I. Pazsit, A. Yamamoto, Y. Yamane,

    Ann. Nucl. Energy   34 巻   頁: 385-395   2007年5月

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    記述言語:英語   掲載種別:研究論文(学術雑誌)  

  306. Erbia-bearing Super High Burnup Fuel: A Pathway for Breaking 5wt% Enrichment Barrier in LWR Fuel 査読有り

    A. Yamamoto, T. Takeda, H. Unesaki, M. Mori,M. Yamasaki

    Proc. International Conference on Nuclear Engineering, ICONE15   CD-ROM 巻   2007年4月

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    DOI: CD-ROM

  307. Performance of the Diffusion and Simplified PN Theories For BWR Pin-By-Pin Fine Mesh Core Analyses 査読有り

    A. Yamamoto K. Tada, Y. Kitamura, Y. Yamane M. Watanabe, H. Noda

    Proc. Joint International Topical Meeting on Mathematics & Computation and Supercomputing in Nuclear Applications (M&C + SNA 2007)   CD-ROM 巻   2007年4月

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    担当区分:筆頭著者   記述言語:英語  

    DOI: CD-ROM

  308. Applicability of The Diffusion and Simplified P3 Theories for BWR Pin-by-pin Core Analysis 査読有り

    K. Tada, A. Yamamoto, M. Watanabe, H. Noda, Y. Kitamura, Y. Yamane

    Proc. International Conference on Nuclear Engineering, ICONE15   CD-ROM 巻   2007年4月

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    記述言語:英語  

    DOI: CD-ROM

  309. Fuel Loading Pattern Optimization Based on Case-Based Reasoning (CBR) 査読有り

    M Adachi, A. Yamamoto, Y. Yamane, Y. Kitamura

    Proc. International Conference on Nuclear Engineering, ICONE15   CD-ROM 巻   2007年4月

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    記述言語:英語  

    DOI: CD-ROM

  310. Evaluation of Core Characteristics of FBR and PWR Loaded with Americium-containing Cermet Fuel 査読有り

    S. Takano, Y. Yamane, A. Yamamoto, M. Osaka, T. Misawa

    Proc. International Conference on Nuclear Engineering, ICONE15   CD-ROM 巻   2007年4月

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    記述言語:英語  

    DOI: CD-ROM

  311. Application of the Mobile-Chord Method for the Method of Characteristics 査読有り

    A. Yamamoto

    Proc. International Conference on Nuclear Engineering, ICONE15   CD-ROM 巻   2007年4月

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    担当区分:筆頭著者   記述言語:英語  

    DOI: CD-ROM

  312. Neutron Transport Models of AEGIS: An Advanced Next-Generation Neutronics Design System 招待有り 査読有り

    N. Sugimura, A. Yamamoto, T. Ushio, M. Mori, M. Tabuchi, T. Endo

    Nucl. Sci. Eng.   155 巻   頁: 276-289   2007年2月

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    記述言語:英語   掲載種別:研究論文(学術雑誌)  

  313. Numerical Solution of Stiff Burnup Equation with Short Half Lived Nuclides by the Krylov Subspace Method 査読有り

    A. Yamamoto, M. Tatsumi, N. Sugimura

    J. Nucl. Sci. Technol.   44 巻   頁: 147-154   2007年2月

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    担当区分:筆頭著者   記述言語:英語   掲載種別:研究論文(学術雑誌)  

  314. *Derivation of Optimum Polar Angle Quadrature Set for the Method of Characteristics Based on Approximation Error of the Bickley Function 査読有り

    A. Yamamoto, M. Tabuchi, N. Sugimura, T. Ushio, M. Mori

    J. Nucl. Sci. Technol.   44 巻   頁: 129-136   2007年2月

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    担当区分:筆頭著者   記述言語:英語   掲載種別:研究論文(学術雑誌)  

  315. Improvement of Spatial Discretization Error on the Semi-analytic Nodal Method using the Scattered Source Subtraction Method 査読有り

    A. Yamamoto, M. Tatsumi

    J. Nucl. Sci. Technol.   43 巻   頁: 1481-1489   2006年12月

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    担当区分:筆頭著者   記述言語:英語   掲載種別:研究論文(学術雑誌)  

  316. Reduction of the Spatial Discretization Error in the Method of Characteristics using the Diamond-difference Scheme 査読有り

    A. Yamamoto

    J. Nucl. Sci. Technol.   43 巻   頁: 1327-1335   2006年11月

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    担当区分:筆頭著者   記述言語:英語   掲載種別:研究論文(学術雑誌)  

  317. Effect of error propagation of nuclide number densities on Monte Carlo burn-up calculations 査読有り

    M. Tohjoh, T. Endo, M. Watanabe, A. Yamamoto

    Ann. Nucl. Energy   33 巻   頁: 1424-1436   2006年11月

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    記述言語:英語   掲載種別:研究論文(学術雑誌)  

  318. Evaluation of Dancoff Factors in Complicated Geometry Using the Method of Characteristics 査読有り

    N. Sugimura, A. Yamamoto

    J. Nucl. Sci. Technol.   43 巻   頁: 1182-1187   2006年10月

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    記述言語:英語   掲載種別:研究論文(学術雑誌)  

  319. Generation of Cross Section Library for Lattice Physics Code, AEGIS 査読有り

    A. Yamamoto, K. Tada, N. Sugimura, T. Ushio, M. Mori

    Proc. Physor-2006   CD-ROM 巻   2006年9月

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    担当区分:筆頭著者   記述言語:英語  

    DOI: DOI

  320. Development of the New Pin-by-Pin Core Calculation Method with Embedded Heterogeneous Assembly Calculation 査読有り

    K. Yamaji, H. Matsumoto, M. Nakano, T. Takeda, A. Yamamoto

    Proc. Physor-2006   CD-ROM 巻   2006年9月

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    記述言語:英語  

    DOI: CD-ROM

  321. Verification of AEGIS/SCOPE2, a Next-Generation In-Core Fuel Management System 査読有り

    M. Tatsumi, N. Sugimura, A. Yamamoto

    Proc. Physor-2006   CD-ROM 巻   2006年9月

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    記述言語:英語  

    DOI: CD-ROM

  322. Calculation Models of AEGIS/SCOPE2, a Core Calculation System of Next Generation 査読有り

    N. Sugimura, T. Ushio, A. Yamamoto, M. Tatsumi

    Proc. Physor-2006   CD-ROM 巻   2006年9月

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    記述言語:英語  

    DOI: CD-ROM

  323. Application of Pin-by-pin Fine Mesh Calculation Method to BWR Core Analysis 査読有り

    K. Tada, A. Yamamoto, Y, Kitamura, Y. Yamane

    Proc. Physor-2006   CD-ROM 巻   2006年9月

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    記述言語:英語  

    DOI: CD-ROM

  324. Feynman-alpha experiment with stationary multiple emission sources 査読有り

    Y. Kitamura, T. Misawa, A. Yamamoto, Y. Yamane, C. Ichihara, H. Nakamura

    Progress in Nuclear Energy   48 巻   頁: 569-577   2006年8月

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    記述言語:英語   掲載種別:研究論文(学術雑誌)  

  325. Application of Variance-to-mean Technique to Subcriticality Monitoring for Accelerator-driven Sub-critical Reactor 査読有り

    Y. Kitamura, K. Taguchi, A. Yamamoto, Y. Yamane, T. Misawa, T. Ichihara, C. Nakamura, H. Oigawa

    Int. J. of Nucl. Energy Sci. and Technol.   2 巻   頁: 266-284   2006年7月

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    記述言語:英語   掲載種別:研究論文(学術雑誌)  

  326. Derivation of Theoretical Formula for the Third Order Neutron Correlation Technique by Using Importance Function 査読有り

    T. Endo, Y. Yamane, A. Yamamoto

    Ann. Nucl. Energy   33 巻   頁: 857-868   2006年7月

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    記述言語:英語   掲載種別:研究論文(学術雑誌)  

  327. Application of the Krylov Subspace Method to Burnup Calculation for Lattice Physics Code 査読有り

    A. Yamamoto, M. Tatsumi, N. Sugimura

    Trans. Am. Nucl. Soc.   95 巻   頁: 713-714   2006年6月

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    担当区分:筆頭著者   記述言語:英語  

  328. Development of Erbia-bearing Super High Burnup Fuel 査読有り

    A. Yamamoto, T. Takeda, H. Unesaki, M. Mori, M. Yamasaki

    Proc. International Congress on Advances in Nuclear Power Plants, ICAPP06     頁: 1874-1882   2006年6月

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    担当区分:筆頭著者   記述言語:英語  

  329. Reduction in Workload of BWR In-Core Fuel Shuffling by New Optimization Methods 査読有り

    A. Yamamoto, M. Tohjoh, K. Komori, Y. Kitamura, Y. Yamane

    Nucl. Technol.   154 巻   頁: 318-327   2006年6月

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    担当区分:筆頭著者   記述言語:英語   掲載種別:研究論文(学術雑誌)  

  330. Improvement on Multi-group Scattering Matrix in Thermal Energy Range Generated by NJOY 査読有り

    A. Yamamoto, N. Sugimura

    Ann. Nucl. Energy   33 巻   頁: 555-559   2006年4月

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    担当区分:筆頭著者   記述言語:英語   掲載種別:研究論文(学術雑誌)  

  331. Space and Energy Dependent Theoretical Formula for the Third Order Neutron Correlation Technique 査読有り

    T. Endo, Y. Yamane, A. Yamamoto

    Ann. Nucl. Energy   33 巻   頁: 521-537   2006年4月

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    記述言語:英語   掲載種別:研究論文(学術雑誌)  

  332. Three-dimensional Pin Power Reconstruction for the Axially Heterogeneous Region in BWR 査読有り

    M. Tohjoh,M. Watanabe,A. Yamamoto

    Ann. Nucl. Energy   33 巻   頁: 242-251   2006年2月

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    記述言語:英語   掲載種別:研究論文(学術雑誌)  

  333. Study of the Spatial Discretization and Temperature Distribution Approximation Effects on BWR Assembly Calculations 査読有り

    M. Tohjoh, M. Watanabe, A. Yamamoto

    Ann. Nucl. Energy   33 巻   頁: 170-179   2006年1月

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    記述言語:英語   掲載種別:研究論文(学術雑誌)  

  334. Calculation of the Stochastic Pulsed Rossi-alpha Formula and its Experimental Verification 査読有り

    Y. Kitamura, K. Taguchi, T. Misawa, I, Pazsit, A. Yamamoto, Y. Yamane, C. Ichihara, H. Nakamura, H. Oigawa

    Prog. Nucl. Energy   48 巻   頁: 37-50   2006年1月

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    記述言語:英語   掲載種別:研究論文(学術雑誌)  

  335. Improvement of the Flat Source Approximation in the Method of Characteristics 査読有り

    A. Yamamoto

    Trans. Am. Nucl. Soc.   95 巻   頁: 577-578   2006年

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    担当区分:筆頭著者   記述言語:英語  

  336. Generalized Coarse Mesh Rebalance Method for Acceleration of Neutron Transport Calculations 査読有り

    A. Yamamoto

    Nucl. Sci. Eng.   151 巻   頁: 274-282   2005年11月

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    担当区分:筆頭著者   記述言語:英語   掲載種別:研究論文(学術雑誌)  

  337. Verification Calculations of AEGIS, an Advanced Neutronics Solver of Next-Generation 査読有り

    N. Sugimura, T. Ushio, A. Yamamoto

    Trans. Am. Nucl. Soc.   92 巻   頁: 633-634   2005年11月

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    記述言語:英語  

  338. Calculation Models of AEGIS, an Advanced Neutronics Solver of Next-Generation 査読有り

    A. Yamamoto, N. Sugimura, T. Ushio

    Trans. Am. Nucl. Soc.   92 巻   頁: 631-632   2005年11月

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    担当区分:筆頭著者   記述言語:英語  

  339. A New Optimization Algorithm for In-core Fuel Shuffling Sequence of BWR 査読有り

    A. Yamamoto, M. Toujou, K. Komori, Y. Kitamura, Y. Yamane

    Trans. Am. Nucl. Soc.   92 巻   頁: 605-606   2005年11月

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    担当区分:筆頭著者   記述言語:英語  

  340. Effects of the spatial discritization and temperature distribution approximation on BWR assembly calculations 査読有り

    M. Tohjoh, M. Watanabe, A. Yamamoto

    Proc. International topical Meeting on Mathematics and Computation, Supercomputing, Reactor Physics and Nuclear and Biological Applications (M&C2005)   CD-ROM 巻   2005年9月

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    記述言語:英語  

    DOI: CD-ROM

  341. Medical Isotope Production Using Pressurized Water Reactor 査読有り

    T. Endo, A. Yamamoto

    Proc. International Symposium on Isotope Science and Engineering from Basic to Applications   CD-ROM 巻   2005年9月

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    記述言語:英語  

    DOI: CD-ROM

  342. Efficient Calculation of Time-Dependent Neutron Transport Equation Using The Constrained Interpolated Profile (CIP) Method 査読有り

    S. Kanoh, T. Endo, A. Yamamoto, Y. Yamane, Y. Kitamura

    Proc. International topical Meeting on Mathematics and Computation, Supercomputing, Reactor Physics and Nuclear and Biological Applications (M&C2005)   CD-ROM 巻   2005年9月

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    記述言語:英語  

    DOI: CD-ROM

  343. Non-equidistant Ray Tracing for the Method of Characteristics 査読有り

    A. Yamamoto, M. Tabuchi, N. Sugimura, T. Ushio

    Proc. International topical Meeting on Mathematics and Computation, Supercomputing, Reactor Physics and Nuclear and Biological Applications (M&C2005)   CD-ROM 巻   2005年9月

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    担当区分:筆頭著者   記述言語:英語  

    DOI: CD-ROM

  344. Development of Advanced Neutronics Design System of Next Generation, AEGIS 査読有り

    N. Sugimura, T. Ushio, M. Mori, A. Yamamoto, M. Tabuchi, T. Endo

    Proc. International topical Meeting on Mathematics and Computation, Supercomputing, Reactor Physics and Nuclear and Biological Applications (M&C2005)   CD-ROM 巻   2005年9月

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    記述言語:英語  

    DOI: CD-ROM

  345. Cost Evaluation of Power Uprate due to Reduction of In-core Power Peaking Factor 査読有り

    M. Adachi, A. Yamamoto, Y. Yamane, Y. Kitamura

    Trans. Am. Nucl. Soc.   93 巻   頁: 378-379   2005年6月

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    記述言語:英語  

  346. Effect of Anisotropic Scattering in PWR/APWR Radial-Reflector Calculations 査読有り

    A. Yamamoto, N. Sugimura, T. Ushio

    Trans. Am. Nucl. Soc.   93 巻   頁: 617-618   2005年6月

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    担当区分:筆頭著者   記述言語:英語  

  347. Yet Another Optimum Polar Angle Quadrature Set for the Method of Characteristics 査読有り

    M. Tabuchi, A. Yamamoto, T. Endo, N. Sugimura, T. Ushio, M. Mori

    Trans. Am. Nucl. Soc.   93 巻   頁: 506-507   2005年6月

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    記述言語:英語  

  348. Application of continuous-energy Monte Carlo code as a cross-section generator of BWR core calculations 査読有り

    M. Tohjoh, M. Watanabe, A. Yamamoto

    Ann. Nucl. Energy   32 巻   頁: 857-875   2005年5月

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    記述言語:英語   掲載種別:研究論文(学術雑誌)  

  349. Calculation of the Pulsed Feynman- and Rossi-alpha Formulae with Delayed Neutrons 査読有り

    Y. Kitamura, I. Pazsit, J. Wright, A. Yamamoto, Y. Yamane

    Ann. Nucl. Energy   32 巻   頁: 671-692   2005年5月

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    記述言語:英語   掲載種別:研究論文(学術雑誌)  

  350. Convergence Property of Response Matrix Method for Various Finite-Difference Formulations used in the Non-Linear Acceleration Method 査読有り

    A. Yamamoto

    Nucl. Sci. Eng.   149 巻   頁: 259-269   2005年3月

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    担当区分:筆頭著者   記述言語:英語   掲載種別:研究論文(学術雑誌)  

  351. A Conceptual Design Study of Proliferation-Resistant PWR Fuel using Reprocessed Uranium 査読有り

    A. Yamamoto, Y. Kagagiri, Y. Yamane

    Nucl. Eng. Des.   235 巻   頁: 649-660   2005年3月

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    担当区分:筆頭著者   記述言語:英語   掲載種別:研究論文(学術雑誌)  

  352. Impact of Pin-by-pin Thermal-hydraulic Feedback Modeling on Steady-state Core Characteristics 査読有り

    A. Yamamoto, T. Ikeno

    Nucl. Technol.   149 巻   頁: 175-188   2005年2月

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    担当区分:筆頭著者   記述言語:英語   掲載種別:研究論文(学術雑誌)  

  353. Improvement of the SPH Method for Pin-by-pin Core Calculations 査読有り

    A. Yamamoto, M. Tatsumi, Y. Kitamura, Y. Yamane

    J. Nucl. Sci. Technol.   41 巻   頁: 1155-1165   2004年12月

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    担当区分:筆頭著者   記述言語:英語   掲載種別:研究論文(学術雑誌)  

  354. Simultaneous Loading Patterns Optimization for Two Successive Cycles of Pressurized Water Reactors 査読有り

    A. Yamamoto, E. Sugimura, Y. Kitamura, Y. Yamane

    J. Nucl. Sci. Technol.   41 巻   頁: 1065-1074   2004年11月

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    担当区分:筆頭著者   記述言語:英語   掲載種別:研究論文(学術雑誌)  

  355. Approximate Treatment of Thermal Expansion Effect in Lattice Transport Calculations 査読有り

    A. Yamamoto, Y. Kitamura, Y. Yamane

    J. Nucl. Sci. Technol.   41 巻   頁: 1003-1007   2004年10月

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    担当区分:筆頭著者   記述言語:英語   掲載種別:研究論文(学術雑誌)  

  356. Convergence Improvement of Coarse Mesh Rebalance Method for Neutron Transport Calculations 査読有り

    A. Yamamoto, Y. Kitamura, T. Ushio, N. Sugimura

    J. Nucl. Sci. Technol.   41 巻   頁: 781-789   2004年8月

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    担当区分:筆頭著者   記述言語:英語   掲載種別:研究論文(学術雑誌)  

  357. Acceleration of Response Matrix Method Using Cross Section Scaling 査読有り

    A. Yamamoto

    Nucl. Sci. Eng.   147 巻   頁: 176-184   2004年6月

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    担当区分:筆頭著者   記述言語:英語   掲載種別:研究論文(学術雑誌)  

  358. Computational Efficiencies of Approximated Exponential Functions for Transport Calculations of the Characteristics Method 査読有り

    A. Yamamoto, Y. Kitamura, Y. Yamane

    Ann. Nucl. Energy   31 巻   頁: 1027-1037   2004年6月

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    担当区分:筆頭著者   記述言語:英語   掲載種別:研究論文(学術雑誌)  

  359. Comparison of 2-Group/9-Group Nodal-Transport Calculations in 3-D Pin-by-Pin Geometry 査読有り

    M. Tatsumi, A. Yamamoto

    Trans. Am. Nucl. Soc.   91 巻   頁: 264-247   2004年6月

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    記述言語:英語  

  360. Comparison of Subcriticality Monitoring Methods for Accelerator-Driven System 査読有り

    K. Taguchi, Y. Yamane, Y. Kitamura, A. Yamamoto

    Trans. Am. Nucl. Soc.   91 巻   頁: 751-752   2004年6月

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    記述言語:英語  

  361. Simultaneous In-core Optimization of PWR Tandem Cycles 査読有り

    A. Yamamoto, E. Sugimura, Y. Kitamura, Y. Yamane

    Trans. Am. Nucl. Soc.   91 巻   頁: 766-767   2004年6月

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    担当区分:筆頭著者   記述言語:英語  

  362. Cell Homogenization Methods for Pin-by-pin Core Calculations Tested in Slab Geometry 査読有り

    A. Yamamoto, Y. Kitamura, Y. Yamane

    Ann. Nucl. Energy   31 巻   頁: 825-847   2004年5月

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    担当区分:筆頭著者   記述言語:英語   掲載種別:研究論文(学術雑誌)  

  363. Improvement of the SPH Method for Multi-assembly Calculations 査読有り

    A. Yamamoto, M. Tatsumi, Y. Kitamura, Y. Yamane,

    Proc. The Physics of Fuel Cycles and Advanced Nuclear Systems: Global Developments (Physor2004)   CD-ROM 巻   2004年4月

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    担当区分:筆頭著者   記述言語:英語  

    DOI: CD-ROM

  364. Derivation of the Space and Energy Dependent Formula for the Third Order Neutron Correlation Technique 査読有り

    T. Endo, Y. Kitamura, A. Yamamoto, Y. Yamane

    Proc. The Physics of Fuel Cycles and Advanced Nuclear Systems: Global Developments (Physor2004)   CD-ROM 巻   2004年4月

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    記述言語:英語  

    DOI: DOI

  365. The feasibility study of the minimum-shuffling reloading strategy for PWR 査読有り

    M. Tabuchi, Y. Hanayama, M. Yamasaki, A. Yamamoto

    Proc. The Physics of Fuel Cycles and Advanced Nuclear Systems: Global Developments (Physor2004)   CD-ROM 巻   2004年4月

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    記述言語:英語  

    DOI: CD-ROM

  366. Sensitivity Analysis for Multiplication Factor Change of LWR Cell Caused by the Differences between JENDL3.2 and JENDL3.3 査読有り

    K. Kitada, T. Takeda, M. Yamasaki, M. tatsumi, A. Yamamoto

    J. Nucl. Sci. Technol.   41 巻   頁: 163-170   2004年2月

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    記述言語:英語   掲載種別:研究論文(学術雑誌)  

  367. A Simple and Efficient Control Rod Cusping Model for Three-Dimensional Pin-By-Pin Core Calculations 査読有り

    A. Yamamoto

    Nucl. Technol.   145 巻   頁: 11-17   2004年1月

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    担当区分:筆頭著者   記述言語:英語   掲載種別:研究論文(学術雑誌)  

  368. Analytic Derivation of the Correction Factor for the Improved Coarse Mesh Method 査読有り

    A. Yamamoto

    Ann. Nucl. Energy   31 巻   頁: 71-86   2004年1月

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    担当区分:筆頭著者   記述言語:英語   掲載種別:研究論文(学術雑誌)  

  369. Convergence Improvements of Response Matrix Method with Large Discontinuity Factors 査読有り

    A. Yamamoto

    Nucl. Sci. Eng.   145 巻   頁: 291-298   2003年11月

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    担当区分:筆頭著者   記述言語:英語   掲載種別:研究論文(学術雑誌)  

  370. Application of Neural Network for Loading Pattern Screening of In-Core Optimization Calculations 査読有り

    A. Yamamoto

    Nucl. Technol.   144 巻   頁: 63-75   2003年10月

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    担当区分:筆頭著者   記述言語:英語   掲載種別:研究論文(学術雑誌)  

  371. Pin-by-pin Thermal-Hydraulic Feedback Modeling in Three-Dimensional Fine-Mesh Core Calculations 査読有り

    A. Yamamoto, T. Ikeno

    Proc. Advances in Nuclear Fuel Management III   CD-ROM 巻   2003年10月

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    担当区分:筆頭著者   記述言語:英語  

    DOI: CD-ROM

  372. MERIT-Factor: A New Concept for Evaluation of the Economic Efficiency of Core Loading Patterns 査読有り

    M. Yamasaki, M. Yoshikuni, A. Yamamoto

    Proc. Advances in Nuclear Fuel Management III   CD-ROM 巻   2003年10月

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    記述言語:英語  

    DOI: CD-ROM

  373. Study on Neutronics Design of Accelerator Driven Subcritical Reactor as Future Neutron Source, Part1; Static Characteristics 査読有り

    A. Yamamoto, S. Shiroya

    Ann. Nucl. Energy   30 巻   頁: 1409-1424   2003年9月

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    担当区分:筆頭著者   記述言語:英語   掲載種別:研究論文(学術雑誌)  

  374. Study on Neutronics Design of Accelerator Driven Subcritical Reactor as Future Neutron Source, Part2; Kinetic Characteristics 査読有り

    A. Yamamoto, S. Shiroya

    Ann. Nucl. Energy   30 巻   頁: 1425-1435   2003年9月

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    担当区分:筆頭著者   記述言語:英語   掲載種別:研究論文(学術雑誌)  

  375. PWR Core Tracking Using a Next-Generation Core Calculation Code, SCOPE2 査読有り

    M. Tatsumi, A. Yamamoto, H. Nagano, K. Sengoku

    Proc. Int.l Conf. on Global Environment and Advanced Nuclear Power Plants (GENES4/ANP2003)   CD-ROM 巻   2003年9月

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    記述言語:英語  

    DOI: CD-ROM

  376. Performance of a Fine-Grained Parallel Model for Multi-Group Nodal-Transport Calculations in Three-Dimensional Pin-By-Pin Reactor Geometry 査読有り

    M. Tatsumi, A. Yamamoto

    Proc. Int. Conf. on Supercomputing in Nuclear Applications (SNA2003)   CD-ROM 巻   2003年9月

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    記述言語:英語  

    DOI: CD-ROM

  377. *Advanced PWR Core Calculation Based on Multi-group Nodal-transport Method in Three-dimensional Pin-by-Pin Geometry 査読有り

    M. Tatsumi, A. Yamamoto

    J. Nucl. Sci. Technol.   40 巻   頁: 376-387   2003年6月

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    記述言語:英語   掲載種別:研究論文(学術雑誌)  

  378. Application of the Distributed Genetic Algorithm for In-Core Fuel Optimization Problems under Parallel Computational Environment 査読有り

    A. Yamamoto, H. Hashimoto

    J. Nucl. Sci. Technol.   39 巻   頁: 1281-1288   2002年12月

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    担当区分:筆頭著者   記述言語:英語   掲載種別:研究論文(学術雑誌)  

  379. Effect of Radial Peaking Factor Limitation on Discharge Burnup 査読有り

    A. Yamamoto

    J. Nucl. Sci. Technol.   39 巻   頁: 1260-1268   2002年12月

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    担当区分:筆頭著者   記述言語:英語   掲載種別:研究論文(学術雑誌)  

  380. Acceleration of Response Matrix Method by Cross Section Scaling 査読有り

    A. Yamamoto

    Proc. International Conference on the New Frontiers of Nuclear Technology: Reactor Physics, Safety and High-Performance Computing     頁: 1A-03   2002年10月

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    担当区分:筆頭著者   記述言語:英語  

  381. SCOPE2:Object-Oriented Parallel Code for Multi-Group Diffusion/Transport Calculations in Three-Dimensional Fine-Mesh Reactor Core Geometry 査読有り

    M. Tatsumi, A. Yamamoto

    Proc. International Conference on the New Frontiers of Nuclear Technology: Reactor Physics, Safety and High-Performance Computing     頁: 12A-01   2002年10月

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    記述言語:英語  

  382. Benchmark Problem Suite for Reactor Physics Study of LWR Next Generation Fuels 査読有り

    A. Yamamoto, T. Ikehara, T. Ito, E. Saji

    J. Nucl. Sci. Technol.   39 巻   頁: 900-912   2002年8月

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    担当区分:筆頭著者   記述言語:英語   掲載種別:研究論文(学術雑誌)  

  383. Object-Oriented Three-Dimensional Fine-Mesh Transport Calculation on Parallel/Distributed Environments for Advanced Reactor Core Analyses 査読有り

    M. Tatsumi, A. Yamamoto

    Nucl. Sci. Eng.   141 巻   頁: 190-217   2002年7月

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    記述言語:英語   掲載種別:研究論文(学術雑誌)  

  384. PWR Core Calculations by the Multigroup Nodal-Transport Method in 3-D Pin-by-Pin Geometry 査読有り

    M. Tatsumi, A. Yamamoto

    Trans. Am. Nucl. Soc.   87 巻   頁: 232-234   2002年6月

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    記述言語:英語  

  385. Basic study on neutronics of future neutron source based on accelerator driven subcritical reactor concept in Kyoto University Research Reactor Institute (KURRI) 査読有り

    S. Shiroya, A. Yamamoto, K. Shin, T. Ikeda, S. Nakano, H. Unesaki

    Prog. Nucl. Energy   40 巻   頁: 489-496   2002年3月

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    記述言語:英語   掲載種別:研究論文(学術雑誌)  

  386. Recent Activities of Loading Pattern Optimization Research in Japan 招待有り 査読有り

    A. Yamamoto

    Trans. Am. Nucl. Soc   84 巻   頁: 57-59   2001年11月

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    担当区分:筆頭著者   記述言語:英語  

  387. A Conceptual Neuronics Design Study for Next Generation Neutron Source in Kyoto University research Reactor Institute (KURRI) 査読有り

    A. Yamamoto, S. Shiroya

    Proc. ICENES 2000: The Tenth International Conference on Emerging Nuclear Energy Systems, Petten   CD-ROM 巻   頁: 66-75   2000年9月

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    担当区分:筆頭著者   記述言語:英語  

  388. Effect of Core Calculation Models on Optimum Cycle Length Analyses of Pressurized Water Reactors 査読有り

    A. Yamamoto, T. Kimoto

    Ann. Nucl. Energy   27 巻   頁: 1039-1050   2000年7月

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    担当区分:筆頭著者   記述言語:英語   掲載種別:研究論文(学術雑誌)  

  389. Analysis of the Ikata-3 Initial Core with the CHAPLET Heterogeneous Transport Calculation Code Based on the Method of Characteristics 招待有り 査読有り

    M. Tatsumi, T.Kimoto, A. Yamamoto

    Trans. Am. Nucl. Soc.   83 巻   頁: 286-287   2000年6月

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    記述言語:英語  

  390. Verification of Cell-Homogenized Whole-Core Transport Calculations in Actual PWR Core Geometry 招待有り 査読有り

    A. Yamamoto, M. Tatsumi, T. Kimoto, S, Kosaka, E. Saji

    Trans. Am. Nucl. Soc.   83 巻   頁: 287-289   2000年6月

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    担当区分:筆頭著者   記述言語:英語  

  391. Effect of Core Calculation Accuracy on Fuel Cycle Cost 査読有り

    A. Yamamoto

    Proc. Advances in Reactor Physics and Mathematics and Computation into the Next Millennium     頁: Ⅶ.A-2   2000年5月

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    担当区分:筆頭著者   記述言語:英語  

  392. A Study on Effects of Pin Cell Homogenization in an Actual Reactor Core Geometry 査読有り

    M. Tatsumi, A. Yamamoto,S.Kosaka,E.Saji

    Proc. Advances in Reactor Physics and Mathematics and Computation into the Next Millennium     頁: IX.D-5   2000年5月

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    記述言語:英語  

  393. A Study on Non-Proliferative, Retrofittable PWR fuel 査読有り

    Y. Katagiri, Y. Yamane, A. Yamamoto

    Proc. of ICONE 8: 8th International Conference on Nuclear Engineering   CD-ROM 巻   頁: 8366   2000年4月

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    記述言語:英語  

  394. Application of Temperature Parallel Simulated Annealing to Loading Pattern Optimizations of Pressurized Water Reactors," 査読有り

    A. Yamamoto, H. Hashimoto

    Nucl Sci. Eng.   136 巻   頁: 247-257   2000年

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    担当区分:筆頭著者   記述言語:英語   掲載種別:研究論文(学術雑誌)  

  395. Loading Pattern Optimizations in a Distributed Parallel Environment 招待有り 査読有り

    A. Yamamoto, H. Hashimoto

    Trans. Am. Nucl. Soc.   80 巻   頁: 223-224   1999年11月

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    担当区分:筆頭著者   記述言語:英語  

  396. Applications of Temperature Parallel Simulated Annealing to Loading Pattern Optimizations of Pressurized Water Reactors 招待有り 査読有り

    A. Yamamoto, H. Hashimoto

    Proc. Mathematics and Computation, Reactor Physics and Environmental Analysis in Nuclear Applications   2 巻   頁: 1445-1458   1999年9月

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    担当区分:筆頭著者   記述言語:英語  

  397. Advanced Reactor Core Analysis by the Object-oriented 3-D Fine Mesh Transport Calculation on Parallel/Distributed Environment 招待有り 査読有り

    M. Tatsumi, A. Yamamoto, H. Hashimoto

    Proc. Mathematics and Computation, Reactor Physics and Environmental Analysis in Nuclear Applications   2 巻   頁: 1288-1297   1999年9月

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    記述言語:英語  

  398. Effective Utilization of Weapon-Grade Plutonium to Upgrade Repeatedly-Reprocessed Mixed-Oxide Fuel for Use in Pressurized Water Reactors 査読有り

    Y. Hanayama, A. Yamamoto K. Kanda

    J. Nucl. Sci. Technol.   36 巻   頁: 746-754   1999年9月

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    記述言語:英語   掲載種別:研究論文(学術雑誌)  

  399. A Comparison Between a One Point Reactor Model And a Two-Dimensional Core Model on Equilibrium Cycle Analysis of Pressurized Water Reactors 査読有り

    A. Yamamoto, T. Kimoto

    Proc. International Conference of the Physics of Nuclear Science and Technology   1 巻   頁: 84-90   1998年10月

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    担当区分:筆頭著者   記述言語:英語  

  400. Object-Oriented Approach for an Iterative Calculation Method and Its Parallelization with Domain Decomposition Method 査読有り

    M. Tatsumi, A. Yamamoto

    Lecture Notes in Computer Science   1505 巻   頁: 1-12   1998年4月

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    記述言語:英語  

  401. Development of Perturbation Code Based on Modified Explicit Higher Order Perturbation Method with Two-Energy-Group 査読有り

    C. H. Pyeon, Y. Yamane, T. Misawa, A. Yamamoto, K. Kagehira, S. Shiroya

    Proc. Joint International Conference on Mathematical Methods and Supercomputing for Nuclear Applications   2 巻   頁: 1149-1158   1997年10月

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    記述言語:英語  

  402. Comparison between Equilibrium Cycle and Successive Multicycle Optimization Methods for In-Core Fuel Management of Pressurized Water Reactors 査読有り

    A. Yamamoto

    Proc. Joint International Conference on Mathematical Methods and Supercomputing for Nuclear Applications   1 巻   頁: 769-781   1997年10月

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    担当区分:筆頭著者   記述言語:英語  

  403. SCOPE:A Scalable and Flexible Parallel Algorithm Based on Object-Oriented Approach for Core Calculations 査読有り

    M. Tatsumi, A. Yamamoto

    Proc. Joint International Conference on Mathematical Methods and Supercomputing for Nuclear Applications   1 巻   頁: 191-202   1997年10月

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    記述言語:英語  

  404. *Comparison between Equilibrium Cycle and Successive Multicycle Optimization Methods for In-Core Fuel Management of Pressurized Water Reactors 査読有り

    A. Yamamoto, K. Kanda

    J. Nucl. Sci. and Technol.   34 巻   頁: 882-892   1997年9月

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    担当区分:筆頭著者   記述言語:英語   掲載種別:研究論文(学術雑誌)  

  405. INSIGHT: An Integrated Scoping Analysis Tool for In-Core Fuel Management of PWR 査読有り

    A. Yamamoto, H. Noda, T. Maruyama, N. Ito

    J. Nucl. Sci. and Technol.   34 巻   頁: 847-855   1997年8月

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    担当区分:筆頭著者   記述言語:英語   掲載種別:研究論文(学術雑誌)  

  406. A Quantitative Comparison of Loading Pattern Optimization Methods for In-Core Fuel Management of PWR 査読有り

    A. Yamamoto

    J. Nucl. Sci. and Technol.   34 巻   頁: 339-347   1997年4月

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    担当区分:筆頭著者   記述言語:英語   掲載種別:研究論文(学術雑誌)  

  407. Resonance Calculations Using the Multiband Method and Interface Current 査読有り

    M. Tatsumi, T. Ito, T. Takeda, M. Yamasaki, A. Yamamoto, M. Takayasu

    Nucl. Sci. Eng.   125 巻   頁: 178-187   1997年3月

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    記述言語:英語   掲載種別:研究論文(学術雑誌)  

  408. INSIGHT: An Integrated Scoping Analysis Tool for In-Core Fuel Management of PWR 査読有り

    A. Yamamoto, H. Noda, N. Ito, T. Maruyama

    Proc. Advances in Nuclear Fuel Management II   1 巻   頁: 8-1-8-11   1997年3月

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    担当区分:筆頭著者   記述言語:英語  

  409. Development of Two-Energy Groups Higher Order Perturbation Method for Application to Core Analysis 査読有り

    C. H. Pyeon, Y. Yamane, T. Misawa, A. Yamamoto, K. Kagehira

    Proc. Advances in Nuclear Fuel Management II   2 巻   頁: 15-11-15-18   1997年3月

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    記述言語:英語  

  410. Core Burnup Calculations Using Neural Networks 査読有り

    H. Noda, A. Yamamoto, Y. Nagasawa, H. Murao, S. Kitamura,

    Proc. Advances in Nuclear Fuel Management II   2 巻   頁: 20-39-20-50   1997年3月

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    記述言語:英語  

  411. Transuranium Fuel Assembly for Transmutation in a Pressurized Water Reactor 招待有り 査読有り

    M. Mori, M. Kawamura, A. Yamamoto

    Nucl. Technol.   117 巻   頁: 171-185   1997年2月

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    記述言語:英語   掲載種別:研究論文(学術雑誌)  

  412. Loading Pattern Optimizations Using Genetic Algorithms 査読有り

    A. Yamamoto

    Proc. International Conference on the Physics of Reactors (PHYSOR96)   3 巻   頁: I-48-I-56   1996年9月

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    担当区分:筆頭著者   記述言語:英語  

  413. Application of Nodal Method to Lambda Mode Higher Harmonics Code 査読有り

    T. Misawa, A. Yamamoto, Y. Yamane

    J. Nucl. Sci. Technol.   33 巻   頁: 668-670   1996年8月

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    記述言語:英語   掲載種別:研究論文(学術雑誌)  

  414. Finite Difference Solution for Multigroup Transport Equation in R-Z Geometry by Spherical Harmonics Method 査読有り

    A. Yamamoto, K. Kobayashi

    J. Nucl. Sci. Technol.   26 巻   頁: 563-574   1989年6月

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    担当区分:筆頭著者   記述言語:英語   掲載種別:研究論文(学術雑誌)  

▼全件表示

書籍等出版物 3

  1. 原子炉の物理

    山本章夫、他( 担当: 共著)

    日本原子力学会  2019年12月  ( ISBN:978-4-89047-172-0

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    記述言語:日本語 著書種別:教科書・概説・概論

  2. Handbook of Nuclear Engineering

    Akio Yamamoto, et al.( 担当: 共著)

    Springer  2010年 

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    記述言語:英語

  3. プルトニウム燃料工学

    山本章夫、他( 担当: 共著)

    1998年 

     詳細を見る

    記述言語:日本語

講演・口頭発表等 82

  1. A New Approach for Resonance Treatment of Doubly Heterogeneous Fuel Using the RSE Method 国際会議

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

    International Conference on Physics of Reactors 2022 (PHYSOR 2022)   2022年5月18日  American Nuclear Society

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    開催年月日: 2022年5月

    記述言語:英語   会議種別:口頭発表(一般)  

    開催地:Pittsburgh   国名:アメリカ合衆国  

  2. Safety assessments of advanced reactors 招待有り 国際会議

    Akio Yamamoto

    Asian Symposium on Risk Assessment and Management ASRAM2021  2021年10月26日 

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    開催年月日: 2021年10月

    記述言語:英語   会議種別:シンポジウム・ワークショップ パネル(指名)  

    開催地:Online  

  3. FRENDY/MG: A Multi-group Cross Section Generation Module using ACE pointwise cross sections 国際会議

    M&C2021  2021年10月5日  American Nuclear Society

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    開催年月日: 2021年10月

    記述言語:英語   会議種別:ポスター発表  

  4. FRENDY/MGの開発 (3)バックグラウンド断面積点の自動生成

    山本章夫

    日本原子力学会 2021年秋の大会  2021年9月10日  日本原子力学会

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    開催年月日: 2021年9月

    記述言語:日本語   会議種別:口頭発表(一般)  

    開催地:オンライン  

  5. 標準委員会の基本方針と今後の戦略について(1)標準委員会の基本方針 招待有り

    山本章夫

    日本原子力学会 2021年秋の大会  2021年9月10日  日本原子力学会

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    開催年月日: 2021年9月

    記述言語:日本語   会議種別:口頭発表(招待・特別)  

    開催地:オンライン  

  6. Verification of the Multi-Group Generation Capability of FRENDY Nuclear Data Processing Codefor Recent Nuclear Data through Comparison of One-group Reaction Rates 国際会議

    2021年6月15日 

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    開催年月日: 2021年6月

    記述言語:英語   会議種別:口頭発表(一般)  

  7. Contaminated Water Management: Current Situation and Issues in Fukushima-Daiichi Nuclear Power Station. 招待有り 国際会議

    The Management of Contaminated Water at Fukushima Daiichi  2021年4月21日  National University of Singapore

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    開催年月日: 2021年4月

    記述言語:英語   会議種別:口頭発表(招待・特別)  

  8. 次期軽水炉における深層防護の実装と技術要件について

    山本章夫

    日本原子力学会 2020年秋の大会 

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    開催年月日: 2020年9月

    記述言語:日本語   会議種別:口頭発表(一般)  

    開催地:オンライン   国名:日本国  

  9. FRENDY/MGの開発 (1)多群断面積作成機能の概要

    山本章夫, 遠藤知弘, 千葉豪, 多田健一

    日本原子力学会 2020年秋の大 

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    開催年月日: 2020年9月

    記述言語:日本語   会議種別:口頭発表(一般)  

    開催地:オンライン   国名:日本国  

  10. 計算科学を活用した炉物理研究の最先端 (1) 原子炉炉心解析と計算科学 招待有り

    山本章夫

    日本機械学会2020年度年次大会 

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    開催年月日: 2020年9月

    記述言語:日本語   会議種別:口頭発表(一般)  

    開催地:オンライン   国名:日本国  

  11. Development of FRENDY nuclear data processing code: Generation capability of multi-group cross sections from ace file 国際会議

    Yamamoto A.

    Transactions of the American Nuclear Society 

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    開催年月日: 2020年6月

    記述言語:英語   会議種別:口頭発表(一般)  

    DOI: 10.13182/T122-32047

    Scopus

  12. Discontinuity Factor; A Discontinuity Condition for Angular Flux? 国際会議

    RPHA2019 

     詳細を見る

    開催年月日: 2019年12月

    記述言語:英語   会議種別:口頭発表(一般)  

    国名:日本国  

  13. Resonance Calculation Using Energy Spectral Expansion Based on Reduced Order Model: Application to Heterogeneous Geometry 国際会議

    ANS 2019 Winter Meeting 

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    開催年月日: 2019年11月

    記述言語:英語   会議種別:口頭発表(一般)  

    国名:アメリカ合衆国  

  14. Uncertainty Quantification/Reduction of BWR Core Characteristics Considering Cross Section and Thermal-Hydraulics Uncertainties 国際会議

    ANS 2019 Annual Meeting 

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    開催年月日: 2019年6月

    記述言語:英語   会議種別:口頭発表(一般)  

    国名:アメリカ合衆国  

  15. Reduced Order Modelに基づくエネルギー展開基底を用いた共鳴計算 (1)理論

    山本章夫、他

    日本原子力学会 2019年 春の年会 

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    開催年月日: 2019年3月

    記述言語:日本語   会議種別:口頭発表(一般)  

    開催地:茨城大学水戸キャンパス   国名:日本国  

  16. Transport Consistent Diffusion Coefficient for CMFD Acceleration 国際会議

    A. Yamamoto, A. Giho, T. Endo

    ANS 2018 Winter Meeting 

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    開催年月日: 2018年11月

    記述言語:英語   会議種別:口頭発表(一般)  

    開催地:Orlando, FL, USA   国名:アメリカ合衆国  

  17. A Simple Treatment of Bowed Assembly Gap Through Correction of Cross Section 国際会議

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

    ANS 2018 Winter Meeting 

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    開催年月日: 2018年11月

    記述言語:英語   会議種別:口頭発表(一般)  

    開催地:Orlando, FL, USA   国名:アメリカ合衆国  

  18. Cache Efficient Flux Region Assignment for the Method of Characteristics 国際会議

    A. Yamamoto, A. Giho, T. Endo

    PHYSOR2018 

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    開催年月日: 2018年4月

    記述言語:英語   会議種別:口頭発表(一般)  

    開催地:Cancun,Mexico   国名:メキシコ合衆国  

  19. Inverse Estimation Methods of Unknown Radioactive Source for Fuel Debris Search 国際会議

    S. Sugaya, T. Endo, A. Yamamoto

    PHYSOR2018 

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    開催年月日: 2018年4月

    記述言語:英語   会議種別:口頭発表(一般)  

    開催地:Cancun,Mexico   国名:メキシコ合衆国  

  20. Estimation of Region-Wise Even-Parity Discontinuity Factor for MOC Through Iterative Procedure 国際会議

    A. Yamamoto, A. Giho, T. Endo

    PHYSOR2018 

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    開催年月日: 2018年4月

    記述言語:英語   会議種別:口頭発表(一般)  

    開催地:Cancun,Mexico   国名:メキシコ合衆国  

  21. MOCによる領域毎even-parity不連続因子の計算

    山本章夫、他

    日本原子力学会 2018年 春の年会 

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    開催年月日: 2018年3月

    記述言語:日本語   会議種別:口頭発表(一般)  

    開催地:大阪大学吹田キャンパス   国名:日本国  

  22. Application of the GENESIS Code to the Kobayashi 3D Benchmark Problem 国際会議

    A. Yamamoto, A. Giho, T. Endo

    ANS 2017 Winter Meeting 

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    開催年月日: 2017年10月 - 2017年11月

    記述言語:英語   会議種別:口頭発表(一般)  

    開催地:Washington, D.C., USA   国名:アメリカ合衆国  

  23. 3次元非均質輸送計算コードGENESISの開発: (3)Kobayashi 3Dベンチマーク問題の解析

    山本章夫、他

    日本原子力学会 2017年秋の大会 

     詳細を見る

    開催年月日: 2017年9月

    記述言語:日本語   会議種別:口頭発表(一般)  

    開催地:北海道大学   国名:日本国  

  24. Development of GENESIS, a Three-dimensional Heterogeneous Transport Code based on the LEAF Method 国際会議

    A. Yamamoto, A. Giho, T. Endo

    M&C 2017 

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    開催年月日: 2017年4月

    記述言語:英語   会議種別:口頭発表(一般)  

    開催地:Jeju, Korea   国名:大韓民国  

  25. Application of Simplified Pn Approximation to Angular Distribution of Neutron Source in MOC Calculations 国際会議

    A. Yamamoto, A. Giho, T. Endo

    ANS 2016 Winter Meeting 

     詳細を見る

    開催年月日: 2016年11月

    記述言語:英語   会議種別:口頭発表(一般)  

    開催地:Las Vegas, Nevada, USA   国名:アメリカ合衆国  

  26. Comparison of the Numerical Stability between CMFD and GCMR with Stabilization Techniques 国際会議

    A. Giho, A. Yamamoto, T. Endo

    ANS 2016 Winter Meeting 

     詳細を見る

    開催年月日: 2016年11月

    記述言語:英語   会議種別:口頭発表(一般)  

    開催地:Las Vegas, Nevada, USA   国名:アメリカ合衆国  

  27. 3次元非均質輸送計算コードGENESISの開発 (2)SPn近似を用いた非等方散乱の取り扱い

    山本章夫、他

    日本原子力学会 2016年 秋の大会 

     詳細を見る

    開催年月日: 2016年9月

    記述言語:日本語   会議種別:口頭発表(一般)  

    開催地:久留米シティプラザ, 福岡県   国名:日本国  

  28. GENESIS - a Transport Solver in Three-Dimensional Heterogeneous Geometry based on the Leaf Method 国際会議

    A. Yamamoto, A. Giho, T. Endo

    PHYSOR2016 

     詳細を見る

    開催年月日: 2016年5月

    記述言語:英語   会議種別:口頭発表(一般)  

    開催地:Sun Valley, Idaho, USA   国名:アメリカ合衆国  

  29. LEAF法に基づく3次元非均質輸送計算コードGENESISの開発

    山本章夫、他

    日本原子力学会 2016年 春の年会 

     詳細を見る

    開催年月日: 2016年3月

    記述言語:日本語   会議種別:口頭発表(一般)  

    開催地:東北大学川内キャンパス   国名:日本国  

  30. Discontinuity Factors for Simplified P3 Theory 国際会議

    A. Yamamoto, T. Sakamoto, T. Endo

    Reactor Physics Asia 2015 (RPHA15) 

     詳細を見る

    開催年月日: 2015年9月

    記述言語:英語   会議種別:口頭発表(一般)  

    開催地:Jeju, Korea   国名:大韓民国  

  31. Reactor Physics Activities in Nagoya University 国際会議

    A. Yamamoto, T. Endo

    Reactor Physics Asia 2015 (RPHA15) 

     詳細を見る

    開催年月日: 2015年9月

    記述言語:英語   会議種別:口頭発表(一般)  

    開催地:Jeju, Korea   国名:大韓民国  

  32. Angular Dependent Transmission Probability Method for Fast Reactor Core Transport Analysis 国際会議

    A. Yamamoto, K. Kirimura, K. Yamaji, S. Kosaka, H. Matsumoto

    ANS 2015 Annual Meeting 

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    開催年月日: 2015年6月

    記述言語:英語   会議種別:口頭発表(一般)  

    開催地:San Antonio, USA   国名:アメリカ合衆国  

  33. ランダムサンプリング法を用いた断面積調整法および感度係数評価 (2)断面積調整法-検証計算

    山本章夫、他

    日本原子力学会 2014年 春の年会 

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    開催年月日: 2014年3月

    記述言語:日本語   会議種別:口頭発表(一般)  

    開催地:東京都市大学   国名:日本国  

  34. Estimation of Self-shielding Effect on Uncertainty of Neutronics Characteristicsusing Random Sampling Method and Continuous-energy Slowing-down Calculation 国際会議

    A. Yamamoto, S. Sato, T. Endo

    ANS 2013 Winter Meeting 

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    開催年月日: 2013年11月

    記述言語:英語   会議種別:口頭発表(一般)  

    開催地:Washington D.C., USA   国名:アメリカ合衆国  

  35. Behavior of Higher Order Fission Source Distribution in Monte-Carlo Calculations 国際会議

    A. Yamamoto, K. Sakata, T. Endo

    ANS 2013 Winter Meeting 

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    開催年月日: 2013年11月

    記述言語:英語   会議種別:口頭発表(一般)  

    開催地:Washington D.C., USA   国名:アメリカ合衆国  

  36. Few-Group Macroscopic Cross Section Adjustment for LWRs Using Random Sampling Technique 国際会議

    A. Yamamoto, S. Kato, T. Endo

    ANS 2013 Annual Meeting 

     詳細を見る

    開催年月日: 2013年6月

    記述言語:英語   会議種別:口頭発表(一般)  

    開催地:Atlanta, GA, USA   国名:アメリカ合衆国  

  37. Explicit Estimation of Higher Order Modes in Fission Source Distribution of Monte-Carlo Calculation 国際会議

    A. Yamamoto, K. Sakata, T. Endo

    MC 2013 

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    開催年月日: 2013年5月

    記述言語:英語   会議種別:口頭発表(一般)  

    開催地:Sun Valley, ID, USA   国名:アメリカ合衆国  

  38. Application of the Multigrid Amplitude Function Method for Time-Dependent Transport Equation Using MOC 国際会議

    K. Tsujita, T. Endo, A. Yamamoto

    MC 2013 

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    開催年月日: 2013年5月

    記述言語:英語   会議種別:口頭発表(一般)  

    開催地:Sun Valley, ID, USA   国名:アメリカ合衆国  

  39. Higher Order Treatment on Temporal Derivative of Angular Flux for Time-Dependent MOC 国際会議

    K. Tsujita, T. Endo, A. Yamamoto, Y. Kamiyama, K. Kirimura

    MC 2013 

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    開催年月日: 2013年5月

    記述言語:英語   会議種別:口頭発表(一般)  

    開催地:Sun Valley, ID, USA   国名:アメリカ合衆国  

  40. MOCを用いた高次モード計算手法の開発

    山本章夫、他

    日本原子力学会 2013年 春の年会 

     詳細を見る

    開催年月日: 2013年3月

    記述言語:日本語   会議種別:口頭発表(一般)  

    開催地:近畿大学東大阪キャンパス   国名:日本国  

  41. 中性子角度分布の時間依存性を厳密に考慮した動特性計算手法; (2) 検証計算

    山本章夫、他

    日本原子力学会 2012年 秋の大会 

     詳細を見る

    開催年月日: 2012年9月

    記述言語:日本語   会議種別:口頭発表(一般)  

    開催地:広島大学東広島キャンパス   国名:日本国  

  42. Analysis of Erbia-Loaded Critical Experiments in KUCA Using AEGIS CrossSection Library 国際会議

    A. Yamamoto, T. Endo, X. Wu

    ANS 2012 Annual Meeting 

     詳細を見る

    開催年月日: 2012年6月

    記述言語:英語   会議種別:口頭発表(一般)  

    開催地:Chicago, Illinois,USA   国名:アメリカ合衆国  

  43. Uncertainty Estimation of Core Safety Parameters Using Cross-Correlationsof Covariance Matrix 国際会議

    A. Yamamoto, Y. Yasue, T. Endo, Y. Kodama, Y. Ohoka, M. Tatsumi

    PHYSOR2012 

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    開催年月日: 2012年4月

    記述言語:英語   会議種別:口頭発表(一般)  

    開催地:Knoxville, USA   国名:アメリカ合衆国  

  44. Multi-Physics Nuclear Reactor Simulator for Advanced Nuclear Engineering Education 国際会議

    A Yamamoto

    PHYSOR2012 

     詳細を見る

    開催年月日: 2012年4月

    記述言語:英語   会議種別:口頭発表(一般)  

    開催地:Knoxville, USA   国名:アメリカ合衆国  

  45. Assembly Discontinuity Factor for Angular Flux in Transport Calculation 国際会議

    A. Yamamoto, T. Endo

    ANS 2011 Winter Meeting 

     詳細を見る

    開催年月日: 2011年10月 - 2011年11月

    記述言語:英語   会議種別:口頭発表(一般)  

    開催地:WashingtonD.C., USA   国名:アメリカ合衆国  

  46. A Derivation of Discontinuity Factor for Angular Flux in Integro-Differential Transport Equation 国際会議

    A. Yamamoto, T. Endo, Y. A. Chao

    ANS 2011 Annual Meeting 

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    開催年月日: 2011年6月

    記述言語:英語   会議種別:口頭発表(一般)  

    開催地:Hollywood, Florida, USA   国名:アメリカ合衆国  

  47. Incorporation of the Two-Term Rational Approximation for the Resonance Calculation with the Tone Method 国際会議

    A. Yamamoto, T. Endo, G. Chiba

    ANS 2010 Winter Meeting 

     詳細を見る

    開催年月日: 2010年11月

    記述言語:英語   会議種別:口頭発表(一般)  

    国名:アメリカ合衆国  

  48. 二項有理式近似を用いたToneの手法の改良

    日本原子力学会2010秋の大会 

     詳細を見る

    開催年月日: 2010年9月

    記述言語:日本語   会議種別:口頭発表(一般)  

    国名:日本国  

  49. Utilization of Discontinuity Factor in Integro-differential Type of Boltzmann Transport Equation 国際会議

    PHYSOR 2010 

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    開催年月日: 2010年5月

    記述言語:英語  

  50. 微積分型輸送方程式における不連続因子の取り扱い

    山本章夫

    日本原子力学会2010春の年会 

     詳細を見る

    開催年月日: 2010年3月

    記述言語:日本語   会議種別:口頭発表(一般)  

    国名:日本国  

  51. Application of Discontinuity Factor for Integro-Differential Transport Equation 国際会議

    ANS 2009 Winter meeting 

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    開催年月日: 2009年11月

    記述言語:英語   会議種別:口頭発表(一般)  

  52. A New Framework of Resonance Calculation Method Based on the Sub-group Method (2); Calculation

    ANS 2009 Annual Meeting 

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    開催年月日: 2009年6月

    記述言語:英語  

  53. A New Framework of Resonance Calculation Method Based on the Sub-group Method (1); Theory 国際会議

    ANS 2009 Annual Meeting 

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    開催年月日: 2009年6月

    記述言語:英語  

  54. Optimum In-Core Power Sharing with Multi-cycle Coupling Effect 国際会議

    ANFM IV 

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    開催年月日: 2009年4月

    記述言語:英語  

  55. 不連続エネルギー群構造の中性子輸送計算に基づく実効断面積計算手法の開発;(3)検証計算

    日本原子力学会 2009年 春の年会 

     詳細を見る

    開催年月日: 2009年3月

    記述言語:日本語  

    国名:日本国  

  56. 大型体系におけるモンテカルロ法の信頼性

    日本原子力学会2009秋の大会 

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    開催年月日: 2009年3月

    記述言語:日本語  

  57. Evaluation of the Background Cross Section for Heterogeneous and Complicated Geometry by the Enhanced Neutron Current Method 国際会議

    ANS 2008 Winter Meeting 

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    開催年月日: 2008年11月

    記述言語:英語  

  58. Projected Predictor-corrector Method for Burnup Calculations of Gd-Bearing Fuel Assemblies 国際会議

    PHYSOR 2008 

     詳細を見る

    開催年月日: 2008年9月

    記述言語:英語  

  59. 複雑形状におけるバックグラウンド断面積の評価法

    日本原子力学会 2008年 秋の大会 

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    開催年月日: 2008年9月

    記述言語:日本語  

    国名:日本国  

  60. 燃焼計算へのProjected Predictor-corrector 法の適用

    日本原子力学会 2008年 春の年会 

     詳細を見る

    開催年月日: 2008年3月

    記述言語:日本語  

    国名:日本国  

  61. Development of Er-SHB fuel: Critical Experiments and Analyses of Homogeneously Erbia-Loaded Cores in KUCA 国際会議

    ANS 2007 Winter Meeting 

     詳細を見る

    開催年月日: 2007年11月

    記述言語:英語  

  62. Accuracy of a Rapid Cell-Heterogeneous Calculation Method for LWR Core Analysis 国際会議

    ANS 2007 Winter Meeting 

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    開催年月日: 2007年11月

    記述言語:英語  

  63. Applicability of the SP3 Nodal Method for BWR Pin-by-pin Core Analysis with Staggered Mesh 国際会議

    ANS 2007 Winter Meeting 

     詳細を見る

    開催年月日: 2007年11月

    記述言語:英語  

  64. Approximate Treatments of Anisotropic Scattering in LWR Analysis 国際会議

    ANS 2007 Annual Meeting 

     詳細を見る

    開催年月日: 2007年6月

    記述言語:英語  

  65. Application of the Mobile-Chord Method for the Method of Characteristics 国際会議

    ICONE15 2007年 

     詳細を見る

    開催年月日: 2007年4月

    記述言語:英語   会議種別:口頭発表(一般)  

  66. Erbia-bearing Super High Burnup Fuel: A Pathway for Breaking 5wt% Enrichment Barrier in LWR Fuel 国際会議

    ICONE15 2007年 

     詳細を見る

    開催年月日: 2007年4月

    記述言語:英語   会議種別:口頭発表(一般)  

    国名:日本国  

  67. Performance of the Diffusion and Simplified PN Theories For BWR Pin-By-Pin Fine Mesh Core Analyses 国際会議

    M&C 2007 

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    開催年月日: 2007年4月

    記述言語:英語  

  68. 軽水炉解析における非等方散乱の取り扱い

     詳細を見る

    開催年月日: 2007年3月

    記述言語:日本語   会議種別:口頭発表(一般)  

    国名:日本国  

  69. A Barrier on the Public Communication of Nuclear Technology - How to Interpret Reactor Kinetics

    Proc. International Symposium on EcoTopia Science 2007 

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    開催年月日: 2007年

    記述言語:英語  

    国名:日本国  

  70. Application of the Krylov Subspace Method to Burnup Calculation in Lattice Physics Code 国際会議

    ANS 2006 Winter Meeting 

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    開催年月日: 2006年11月

    記述言語:英語   会議種別:口頭発表(一般)  

  71. Improvement of the Flat Source Approximation in the Method of Characteristics 国際会議

    ANS 2006 Winter Meeting 

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    開催年月日: 2006年11月

    記述言語:英語   会議種別:口頭発表(一般)  

  72. ダイヤモンド差分式を用いたCharacteristics法の精度向上

    日本原子力学会 2006年 秋の大会 

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    開催年月日: 2006年9月

    記述言語:日本語   会議種別:口頭発表(一般)  

    国名:日本国  

  73. Generation of Cross Section Library for Lattice Physics Code, AEGIS 国際会議

    PHYSOR2006 

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    開催年月日: 2006年9月

    記述言語:英語   会議種別:口頭発表(一般)  

  74. Development of Erbia-bearing Super High Burnup Fuel 国際会議

    ICAPP06 

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    開催年月日: 2006年6月

    記述言語:英語   会議種別:口頭発表(一般)  

  75. 次世代炉心計算システムAEGIS/SCOPE2;2ライブラリ

    日本原子力学会 2006年 春の年会 

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    開催年月日: 2006年3月

    記述言語:日本語   会議種別:口頭発表(一般)  

    国名:日本国  

  76. Effect of Anisotropic Scattering in PWR/APWR Radial-Reflector Calculations 国際会議

    ANS 2005 Winter Meeting 

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    開催年月日: 2005年11月

    記述言語:英語   会議種別:口頭発表(一般)  

  77. Non-equidistant Ray Tracing for the Method of Characteristics 国際会議

    M&C 2005 

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    開催年月日: 2005年9月

    記述言語:英語   会議種別:口頭発表(一般)  

  78. Calculation Models of AEGIS, and Advanced Neutronics Solver of Next-Generation 国際会議

    ANS 2005 Annual Meeting 

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    開催年月日: 2005年6月

    記述言語:英語   会議種別:口頭発表(一般)  

  79. A New Optimization Algorithm for In-core Fuel Shuffling Sequence of BWR 国際会議

    ANS 2005 Annual Meeting 

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    開催年月日: 2005年6月

    記述言語:英語   会議種別:口頭発表(一般)  

  80. 次世代非均質輸送計算システムAEGISの開発

    日本原子力学会 2005年 春の年会 

     詳細を見る

    開催年月日: 2005年

    記述言語:日本語   会議種別:口頭発表(一般)  

    国名:日本国  

  81. 非構造メッシュにおける効率的なCharacteristics法の加速(1) 加速法の概要と収束性評価

    日本原子力学会 2004年 春の年会 

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    開催年月日: 2004年

    記述言語:日本語   会議種別:口頭発表(一般)  

    国名:日本国  

  82. 燃料装荷パターンの2サイクル同時最適化

    日本原子力学会 2004年 秋の大会 

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    開催年月日: 2004年

    記述言語:日本語   会議種別:口頭発表(一般)  

    国名:日本国  

▼全件表示

科研費 4

  1. 革新炉の解析精度向上:有効部分空間法を用いた高精度かつロバストな断面積調整

    研究課題/研究課題番号:21K04940  2021年4月 - 2024年3月

    日本学術振興会  科学研究費助成事業(学術研究助成基金助成金)  基盤研究(C)(一般)

    遠藤知弘、丸山修平

      詳細を見る

    担当区分:研究代表者  資金種別:競争的資金

    配分額:3900000円 ( 直接経費:3000000円 、 間接経費:900000円 )

  2. 原子炉設計拡張状態の予測不確かさ定量化:誤差相関を用いた新たな評価手法の開発

    研究課題/研究課題番号:16K06956  2016年4月 - 2019年3月

    科学研究費補助金  

    山本 章夫

      詳細を見る

    担当区分:研究代表者 

    配分額:4680000円 ( 直接経費:3600000円 、 間接経費:1080000円 )

    原子炉の設計基準事故を超える過酷事故条件における核特性シミュレーションの計算誤差を、誤差相関およびランダムサンプリング法を用いて定量化する新たな評価手法を開発した。主成分分析及び地球統計学で用いられるクリギング法を用いて、シミュレーションに現れる各種パラメータと計算誤差の相関を学習し、計算誤差を予測する。
    加圧水型軽水炉で用いられる燃料集合体体系において、通常運転から過酷事故条件までを含む幅広い状態における計算誤差を本手法で予測した。計算誤差は、決定論的手法と連続エネルギーモンテカルロコードの実効増倍率の差異とした。その結果、精度良く計算誤差を予測可能であることを確認した。
    原子炉の安全性は、解析計算により確認される。従って、解析計算の信頼性を確認しておくことは、極めて重要である。原子炉の通常運転時については、実機の測定データが多数得られており、これを用いて解析計算の精度や信頼性を確認することが出来る。一方、過酷事故条件は特殊な実験装置を用いなければ再現が困難であり、解析計算の信頼性を確認できるデータは限られる。
    本研究では、限られた検証データを基に、過酷事故条件を含む幅広い状態で解析計算の計算誤差(予測精度)を評価する手法を開発した。これは、解析計算の信頼性、ひいては、原子炉の安全性確保に寄与する。

  3. 実測困難な原子炉安全パラメータの不確かさ評価-分散共分散行列を用いた新概念

    2012年4月 - 2015年3月

    科学研究費補助金  基盤研究(C)

    山本章夫

      詳細を見る

    担当区分:研究代表者 

    原子炉の炉心解析プログラムで得られる解析結果には、断面積の不確かさや計算手法に起因する誤差が必ず含まれている。予測誤差は、実測可能な安全パラメータに対しては容易に確認できるが、一方で実測が困難な安全
    パラメータも存在する。このような安全パラメータの誤差を確認することは、原子炉の安全性を担保する上で極めて重要である。
     本研究では、実測が困難な原子炉の核的安全性パラメータの予測誤差を評価する理論を新たに確立し、その適用性を確認する。本研究では、①これまで着目されてこなかった安全パラメータ間の誤差の相関(共分散)を評価する理論を確立し、②これを活用することで実測が困難な安全パラメータの予測誤差を推定することを可能とする。本研究の成果により、原子炉の安全評価手法の信頼性を向上させることが可能となる。

  4. 逆解析を用いた燃料配置最適化方法に関する研究

    2007年

    科学研究費補助金  特別研究員奨励費,課題番号:70008084

    山本 章夫

      詳細を見る

    担当区分:研究代表者 

 

担当経験のある科目 (本学) 39

  1. 原子力工学設計演習

    2020

  2. 原子力安全工学

    2020

  3. 安全・信頼性工学

    2020

  4. 数学1及び演習

    2020

  5. 基礎セミナーA

    2020

  6. 原子力安全工学

    2019

  7. 原子炉物理学

    2019

  8. 基礎セミナーA

    2019

  9. 安全・信頼性工学

    2019

  10. 基礎セミナーA

    2018

  11. 原子力安全工学

    2018

  12. 原子炉物理学

    2018

  13. 安全・信頼性工学

    2018

  14. 基礎セミナーA

    2017

  15. 原子力安全工学

    2017

  16. 原子炉物理学

    2017

  17. 安全・信頼性工学

    2017

  18. 基礎セミナーA

    2016

  19. 原子炉物理学

    2016

  20. エネルギー量子制御工学特論

    2016

  21. エネルギー量子制御工学特論

    2015

  22. 基礎セミナーA

    2015

  23. 原子炉物理学

    2015

  24. エネルギーと環境

    2014

  25. 基礎セミナーA

    2014

  26. 原子炉物理学

    2014

  27. エネルギー量子制御工学特論

    2014

  28. エネルギーと環境

    2013

  29. 基礎セミナーA

    2013

  30. 原子炉物理学

    2013

  31. エネルギー量子制御工学特論

    2013

  32. 基礎セミナーB

    2012

  33. 原子炉物理学

    2012

  34. エネルギーと環境

    2012

  35. エネルギー量子制御工学特論

    2012

  36. エネルギーと環境

    2011

  37. 基礎セミナーB

    2011

  38. 原子炉物理学

    2011

  39. エネルギー量子制御工学特論

    2011

▼全件表示