Updated on 2024/05/15

写真a

 
YAMAMOTO, Akio
 
Organization
Graduate School of Engineering Applied Energy 3 Professor
Graduate School
Graduate School of Engineering
Undergraduate School
School of Engineering Energy Science and Engineering
Title
Professor

Degree 1

  1. Ph. D. ( 1999.4   Kyoto University ) 

Research Interests 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

Research Areas 1

  1. Energy Engineering / Nuclear engineering  / Reactor physics, nuclear Safety

Current Research Project and SDGs 2

  1. Advanced reactor simulation methods

  2. Safety analysis of nuclear reactors

Research History 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

    2017.4

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    Country:Japan

  2. 名古屋大学工学研究科マテリアル理工学専攻・教授

    2010.4 - 2017.3

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    Country:Japan

  3. 名古屋大学工学研究科マテリアル理工学専攻・准教授

    2007.4 - 2010.3

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    Country:Japan

  4. Nagoya University   Head

    2004.4 - 2010.3

  5. 名古屋大学工学研究科マテリアル理工学専攻・助教授

    2004.4 - 2007.4

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    Country:Japan

  6. 名古屋大学工学研究科原子核工学専攻・助教授

    2003.4 - 2004.3

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    Country:Japan

  7. Nuclear Fuel Industries, Ltd.

    1989.4 - 2003.3

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    Country:Japan

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Education 3

  1. Kyoto University

    1996.4 - 1998.3

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    Country: Japan

  2. Kyoto University   Graduate School, Division of Engineering   Department of Nuclear Engineering

    1987.4 - 1989.3

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    Country: Japan

  3. Kyoto University   Faculty of Engineering   Department of Nuclear Engineering

    1983.4 - 1987.3

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    Country: Japan

Professional Memberships 2

  1. Atomic Energy Society of Japan

  2. American Nuclear Society

Committee Memberships 1

  1. 文部科学省   国際原子力人材育成イニシアティブ プログラムディレクター  

       

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    Committee type:Government

Awards 14

  1. 日本原子力学会賞 論文賞

    2024.3   日本原子力学会  

    丸山修平、遠藤知弘、山本章夫

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    Award type:Award from Japanese society, conference, symposium, etc.  Country:Japan

  2. 日本原子力学会賞 特賞・技術賞

    2022.3   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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    Award type:Award from Japanese society, conference, symposium, etc.  Country:Japan

  3. 日本原子力学会 計算科学技術部会 功績賞

    2021.3   日本原子力学会 計算科学技術部会  

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    Award type:Award from Japanese society, conference, symposium, etc.  Country:Japan

  4. 日本原子力学会賞 論文賞

    2019.3   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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    Award type:Award from Japanese society, conference, symposium, etc.  Country:Japan

  5. 日本原子力学会 計算科学技術部会 業績賞

    2019.3   日本原子力学会 計算科学技術部会  

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    Award type:Award from Japanese society, conference, symposium, etc.  Country:Japan

  6. 日本原子力学会賞 論文賞

    2018.3   日本原子力学会  

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    Award type:Award from Japanese society, conference, symposium, etc.  Country:Japan

  7. 日本原子力学会賞 論文賞

    2017.3   日本原子力学会  

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    Award type:Award from Japanese society, conference, symposium, etc.  Country:Japan

  8. Fellow, American Nuclear Society

    2016.11   American Nuclear Society  

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    Award type:Award from international society, conference, symposium, etc.  Country:United States

  9. 日本原子力学会賞 論文賞

    2014.3   日本原子力学会  

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    Award type:Award from Japanese society, conference, symposium, etc.  Country:Japan

  10. 米国原子力学会炉物理部会最優秀論文賞

    2011.11   米国原子力学会  

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    Country:United States

  11. 日本原子力学会論文賞

    2010.3   日本原子力学会  

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    Country:Japan

  12. Best Paper Award from the Reactor Physics Division (RPD) of the American Nuclear Society

    2009.6   Reactor Physics Division (RPD) of the American Nuclear Society  

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    Country:United States

  13. 日本原子力学会賞技術賞

    2004.3   日本原子力学会  

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    Country:Japan

  14. 日本原子力学会賞論文賞

    1999.3   日本原子力学会  

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    Country:Japan

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Papers 446

  1. Application of neutron current method for Dancoff factor estimation of fuel particles in double-heterogeneous fuel

    Yamamoto A., Endo T.

    Journal of Nuclear Science and Technology   Vol. 61 ( 3 ) page: 354 - 362   2024

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    Language:English   Publishing type:Research paper (scientific journal)   Publisher:Journal of Nuclear Science and Technology  

    An evaluation method of Dancoff factors of fuel particles in a typical fuel element of a high-temperature gas-cooled reactor (HTGR) is proposed based on the neutron current method that is widely used for lattice physics calculations of light water reactors. The first level of heterogeneity, i.e. dispersed fuel particles in graphite matrix, is treated by the deterministic statistical geometry (DSTG) method. The homogenized cross-sections of dispersed fuel particles are provided to the second level of heterogeneity, i.e. fuel compacts in a fuel element or a fuel pebble. The validity of the present method is confirmed through the comparison with the reference Dancoff factor obtained by the Monte Carlo and the neutron current methods, which explicitly treats the double heterogeneity. The comparison is carried out for a fuel element simulating the High Temperature Engineering Test Reactor (HTTR), which adopts a typical prismatic fuel element. The numerical results indicate that the present method well reproduces the reference Dancoff factor under various packing fractions. Since the present method can handle flexible geometry and its computation time is short, the present method will be a candidate for the Dancoff factor evaluation method in design calculations of HTGR.

    DOI: 10.1080/00223131.2023.2231462

    Web of Science

    Scopus

  2. Deterministic Transport Calculation Method for Statistical Geometry with Small Fuel Particles

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

    Nuclear Science and Engineering   Vol. 198 ( 5 ) page: 981 - 992   2024

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    Language:English   Publishing type:Research paper (scientific journal)   Publisher:Nuclear Science and Engineering  

    A deterministic transport calculation method is proposed for the treatment of dispersed fuel particles in a fuel compact/fuel pebble of a typical high-temperature gas-cooled reactor fuel. The random distribution of fuel particles was considered using the statistical geometry (STG) method, which is widely used in the Monte Carlo method. A long-ray trace, which represents a neutron flight path, was considered, and the segment lengths and material distributions on the ray trace were randomly sampled using STG. Then a conventional transport sweep, as used in the method of characteristics, was performed along the ray trace. The proposed deterministic statistical geometry (DSTG) method can calculate the flux spatial distribution in a heterogeneous geometry containing randomly dispersed fuel particles and the surrounding graphite matrix, which is consistent with the STG in a Monte Carlo method. The validity of the DSTG method was confirmed through sensitivity calculations and comparisons with a multigroup Monte Carlo method that utilizes STG. The proposed method can be used for the homogenization of heterogeneous structures inside a fuel compact or fuel pebble as an alternative to conventional deterministic unit cell calculations that consider fuel particles and the surrounding matrix in high-temperature gas-cooled reactor fuels.

    DOI: 10.1080/00295639.2023.2230414

    Web of Science

    Scopus

  3. Deterministic sampling method using simplex ensemble and scaling method for efficient and robust uncertainty quantification

    Endo T., Maruyama S., Yamamoto A.

    Journal of Nuclear Science and Technology   Vol. 61 ( 3 ) page: 363 - 374   2024

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    Language:English   Publishing type:Research paper (scientific journal)   Publisher:Journal of Nuclear Science and Technology  

    Uncertainty quantification (UQ) of the neutron multiplication factor is important to investigate the appropriate safety margin for a target system. Although the random sampling method is a practical and useful UQ method, a large computational cost is required to reduce the statistical error of the estimated uncertainty. Furthermore, if an input variable follows a normal distribution with a large standard deviation, the perturbed input variable by the random sampling method may become a physically inappropriate or negative value. To address these issues for the efficient and robust UQ, a modified deterministic sampling method using the simplex ensemble and the scaling method is proposed. The features of the proposed method are summarized as follows: The sample size is (Formula presented.), where (Formula presented.) corresponds to the effective rank of the covariance matrix between the input variables; depending on a situation of target UQ, the amounts of perturbations for the input parameters can be arbitrarily given by the scaling factor method; the scaling factor can be updated to avoid physically inappropriate in the perturbed input variables. The effectiveness of the proposed method is demonstrated through the UQ of the neutron multiplication factor due to fuel manufacturing uncertainties for a typical PWR pin-cell burnup calculation.

    DOI: 10.1080/00223131.2023.2231931

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    Scopus

  4. Uncertainty reduction of sodium void reactivity using data from a sodium shielding experiment

    Maruyama S., Endo T., Yamamoto A.

    Journal of Nuclear Science and Technology   Vol. 61 ( 1 ) page: 31 - 43   2024

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    Language:English   Publishing type:Research paper (scientific journal)   Publisher:Journal of Nuclear Science and Technology  

    This study investigated the feasibility of reducing the uncertainty associated with fast-reactor-core design by sharing an experimental database between different fields (e.g. reactor physics and radiation shielding) using data assimilation techniques. As the first step in this study, we focused on the ORNL sodium shielding experiment and investigated the possibility of using the experimental data to reduce the uncertainty in sodium void reactivity (SVR), which is the most important safety parameter for sodium-cooled fast reactors. A sensitivity analysis based on the Generalized Perturbation Theory was performed for the sodium shielding experiment. Using the sensitivity coefficients evaluated here and those of the sodium void reactivity previously evaluated by the JAEA, we showed that sodium shielding experimental data can contribute to the uncertainty reduction of SVR by adopting the cross-section adjustment method. Based on this study, the uncertainty reduction effect is expected to be significant, especially for SVR dominated by neutron-leakage phenomena. Although new reactor physics experimental data on SVR may be difficult to obtain, the results of this study suggest that data from sodium shielding experiments can partially substitute for this role. This study demonstrated the value of the mutual use of integral experimental data in fast reactor designs.

    DOI: 10.1080/00223131.2023.2244512

    Scopus

  5. Impact of Uncertainty Reduction on Lead-Bismuth Coolant in Accelerator-Driven System Using Sample Reactivity Experiments

    Katano R., Oizumi A., Fukushima M., Pyeon C.H., Yamamoto A., Endo T.

    Nuclear Science and Engineering   Vol. 198 ( 6 ) page: 1215 - 1234   2024

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    Language:English   Publishing type:Research paper (scientific journal)   Publisher:Nuclear Science and Engineering  

    In this study, we have demonstrated that data assimilation (DA) using lead and bismuth sample reactivities measured in the Kyoto University Critical Assembly A-core can successfully reduce the uncertainty of the coolant void reactivity in accelerator-driven systems (ADSs) derived from inelastic scattering cross sections of lead and bismuth. We reevaluated and highlighted the experimental uncertainties and correlations of the sample reactivities for the DA formula. We used the MCNP6.2 code to evaluate the sample reactivities and their uncertainties and performed DA using the reactor analysis code system MARBLE. The high-sensitivity coefficients of the sample reactivities to lead and bismuth allowed us to reduce the cross-section–induced uncertainty of the void reactivity of the ADS from 6.3% to 4.8%, achieving a provisional target accuracy of 5% in this study. Furthermore, we demonstrated that the uncertainties arising from other dominant factors, such as minor actinides and steel, can be effectively reduced by using integral experimental data sets for the unified cross-section dataset ADJ2017.

    DOI: 10.1080/00295639.2023.2246779

    Scopus

  6. Report on the lecture of standard committee in the 2023 fall meeting entitled “Standardization Activities for Safe Long Term Operation”

    Murakami Kenta, Onizawa Kunio, Yamamoto Akio

    Journal of the Atomic Energy Society of Japan   Vol. 66 ( 4 ) page: 199 - 202   2024

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    Language:Japanese   Publishing type:Research paper (scientific journal)   Publisher:Atomic Energy Society of Japan  

    DOI: 10.3327/jaesjb.66.4_199

    CiNii Research

  7. Comparison of internal boundary conditions for optical diffusion calculations considering reflection and refraction

    Amano T., Endo T., Yamamoto A.

    OSA Continuum   Vol. 2 ( 7 ) page: 1540 - 1560   2023.6

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    Language:English   Publishing type:Research paper (scientific journal)   Publisher:OSA Continuum  

    A method to treat the internal boundary condition in an optical diffusion calculation is proposed and is compared with the conventional methods. One of the existing internal boundary conditions is Haskel’s method, which uses the effective reflection coefficient for partial currents. However, Haskel’s method ignores incoming partial currents from the adjacent mesh in its derivation. As a result, the accuracy at the internal boundary is lower. This paper proposes a method to improve the accuracy by iteratively updating the effective reflection coefficient for partial current. The proposed method was applied to the benchmark calculations on a one-dimensional slab geometry and its accuracy was confirmed by comparing it with the reference solution obtained by the Monte Carlo code MCML, along with the previously proposed Haskel’s method and Aronson’s method. As a result, it was confirmed that the proposed method is more accurate than Haskel’s method at the internal boundary and gives the same result as Aronson’s method. The convergence of the effective reflection coefficient using iterative calculations in the proposed method was good.

    DOI: 10.1364/OPTCON.492445

    Scopus

  8. Impact of nuclear data revised from JENDL-4.0 to JENDL-5 on PWR spent fuel nuclide composition

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

    Journal of Nuclear Science and Technology   Vol. 60 ( 11 ) page: 1386 - 1396   2023

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    Language:English   Publishing type:Research paper (scientific journal)   Publisher:Journal of Nuclear Science and Technology  

    The burnup calculations for estimating the nuclide composition of the spent fuel are highly dependent on nuclear data. Many nuclides in the latest version of the Japanese Evaluated Nuclear Data Library JENDL-5 were modified from JENDL-4.0 and the modification affects the burnup calculations. This study confirmed the validity of JENDL-5 in the burnup calculations. The PIE data of Takahama-3 was used for the validation. The effect of modifications of the parameters, e.g. cross sections and fission yields, from JENDL-4.0 to JENDL-5 on the nuclide compositions was quantitatively investigated. The calculation results showed that JENDL-5 has a similar performance to JENDL-4.0. The calculation results also revealed that the modifications of the cross sections of actinide nuclides, fission yields, and thermal scattering low data of hydrogen in H2O affected the nuclide compositions of PWR spent fuels.

    DOI: 10.1080/00223131.2023.2201603

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  9. An estimation method for an unknown covariance in cross-section adjustment based on unbiased and consistent estimator

    Maruyama S., Endo T., Yamamoto A.

    Journal of Nuclear Science and Technology   Vol. 60 ( 11 ) page: 1372 - 1385   2023

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    Language:English   Publishing type:Research paper (scientific journal)   Publisher:Journal of Nuclear Science and Technology  

    A new estimation method of an unknown covariance in the cross-section adjustment method for the development of an application library is proposed. The unknown covariance is defined by the difference between the true covariance (the population covariance) and a prior covariance assumed by an analyst. The unknown covariance is estimated using an empirical covariance consistent with the observed data. To estimate the unknown covariance, an unbiased and consistent estimator in regression analysis has been incorporated into the conventional cross-section adjustment. This estimator does not require assumptions for the probability distribution of the observation data. The statistical properties of this estimator were numerically verified. In addition, the effectiveness of the proposed method was confirmed by another numerical test using actual integral experimental data. In the second numerical test, the modeling uncertainty (covariance) due to the deterministic analysis method was assumed to be unknown. The results show that the proposed method can practically estimate the unknown covariance and adjusted cross-sections using only prior information on covariances.

    DOI: 10.1080/00223131.2023.2203707

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

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

    Nuclear Science and Engineering   Vol. 197 ( 2 ) page: 176 - 188   2023

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    Language:English   Publishing type:Research paper (scientific journal)   Publisher: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

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

    Chiba G., Yamamoto A., Tada K.

    Journal of Nuclear Science and Technology   Vol. 60 ( 2 ) page: 132 - 139   2023

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    Language:English   Publishing type:Research paper (scientific journal)   Publisher:Journal of Nuclear Science and Technology  

    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

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  12. Development of ACE file perturbation tool using FRENDY

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

    Journal of Nuclear Science and Technology   Vol. 60 ( 6 ) page: 624 - 631   2023

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    Language:English   Publishing type:Research paper (scientific journal)   Publisher:Journal of Nuclear Science and Technology  

    The sensitivity analysis and the uncertainty quantification play an important role in improving the evaluated nuclear data library. The current computational performance enables us to perform sensitivity analysis and uncertainty quantification using continuous energy Monte Carlo calculation codes. The ACE file perturbation tool was developed for these calculations using modules of FRENDY. This tool perturbs the microscopic cross section, the number of neutrons per fission, and the fission spectrum. The uncertainty quantification using the random sampling method is also available if the user prepares the covariance matrices. The uncertainty of keff using the ACE file perturbation tool was compared to the current sensitivity analysis codes SCALE/TSUNAMI and MCNP/KSEN. The comparison results indicated that the random sampling method using this tool accurately estimates the uncertainty of keff. The ACE file perturbation tool is included in the current version of FRENDY.

    DOI: 10.1080/00223131.2022.2130463

    Web of Science

    Scopus

  13. Nuclear data adjustment using a deterministic sampling method with unscented transformation

    Fukui Y., Endo T., Yamamoto A.

    Journal of Nuclear Science and Technology   Vol. 60 ( 3 ) page: 238 - 250   2023

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    Language:English   Publishing type:Research paper (scientific journal)   Publisher:Journal of Nuclear Science and Technology  

    A nuclear data adjustment method using a deterministic sampling method based on an unscented transform was developed, and its validity was confirmed through a twin experiment using a critical benchmark problem. Conventional nuclear data adjustment methods require sensitivity analysis using generalized perturbation theory or many forward calculations using stochastically sampled nuclear data. To address this issue, this study focused on unscented transform-based sampling (UTS), which is used in uncertainty quantification. Based on the UTS, perturbed nuclear data can be deterministically sampled to reproduce the population covariance matrix with minimum sample size. Therefore, UTS can significantly reduce the computational cost compared to conventional nuclear data adjustment using random sampling (RS). Furthermore, the UTS was improved to prevent the sampling of negative nuclear data while accurately reproducing the population covariance matrix. The proposed method was applied to the numerical experiment of Godiva, and the adjusted nuclear data were compared with those obtained using conventional methods. Consequently, it was demonstrated that UTS can adjust nuclear data at a lower computational cost than RS.

    DOI: 10.1080/00223131.2022.2095051

    Web of Science

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  14. Efficient reduced order model based on the proper orthogonal decomposition for time-dependent MOC calculations

    Tsujita K., Endo T., Yamamoto A.

    Journal of Nuclear Science and Technology   Vol. 60 ( 3 ) page: 343 - 357   2023

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    Language:English   Publishing type:Research paper (scientific journal)   Publisher:Journal of Nuclear Science and Technology  

    An efficient reduced order model (ROM) for time-dependent transport calculations using the method of characteristics (MOC) is proposed. In the present ROM, the flux distributions and the net neutron currents between the adjacent unstructured mesh regions are taken from the MOC solution. Then, the coefficient matrices for the MOC-equivalent diffusion equation are reconstructed from them. The proper orthogonal decomposition (POD) is applied for the MOC-equivalent diffusion equation to reduce the degree of freedoms (DOFs) using the orthogonal bases obtained by the singular value decomposition (SVD) for the sampled MOC solution. The accuracy and computation time of the present ROM are verified in the C5G7-TD 2D benchmark problem. The calculation results show that the present ROM enables us approximately 5000–6000 times faster computation than the full order model (FOM) for kinetic calculations itself in the present calculation condition. The present method can be substituted as real-time simulations without the spatial homogenization when typical flux distributions and the net neutron currents of a target problem can be precalculated.

    DOI: 10.1080/00223131.2022.2097963

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  15. Development of nuclear data processing code FRENDY version 2

    Tada K., Yamamoto A., Kunieda S., Konno C., Kondo R., Endo T., Chiba G., Ono M., Tojo M.

    Journal of Nuclear Science and Technology     2023

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    Language:English   Publishing type:Research paper (scientific journal)   Publisher:Journal of Nuclear Science and Technology  

    Nuclear data processing is important to connect evaluated nuclear data libraries and radiation transport codes. The nuclear data processing code FRENDY version 1 was released in 2019 to generate ACE-formatted cross section files with simple input data. After we released FRENDY version 1, many functions were developed, e.g. neutron multi-group cross section generation, explicit consideration of the resonance interference effect among different nuclides in a material, adaptive setting of the background cross sections, consideration of the resonance upscattering, ACE file perturbation, statistical uncertainty quantification of probability table, and modification of ENDF-6-formatted file. FRENDY version 2 was released including these new functions. It generates GENDF- and MATXS-formatted neutron multi-group cross section files from an ACE-formatted cross section file or an evaluated nuclear data file. This paper explains the features of the new functions implemented in FRENDY version 2 and the verification of the neutron multigroup cross section generation function of this code.

    DOI: 10.1080/00223131.2023.2278600

    Scopus

  16. Application of Equivalent Dancoff Factor Method for Resonance Calculation of Double Heterogeneous Fuel

    Yamamoto A., Endo T.

    Transactions of the American Nuclear Society   Vol. 128   page: 690 - 693   2023

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    Language:English   Publishing type:Research paper (scientific journal)   Publisher:Transactions of the American Nuclear Society  

    DOI: 10.13182/T128-41580

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  17. Sensitivity analysis of risk assessment for continuous Markov process Monte Carlo method using correlated sampling method

    Morishita Y., Yamamoto A., Endo T.

    Journal of Nuclear Science and Technology   Vol. 60 ( 12 ) page: 1573 - 1585   2023

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    Language:English   Publishing type:Research paper (scientific journal)   Publisher:Journal of Nuclear Science and Technology  

    The correlated sampling method is applied to the continuous Markov-chain Monte-Carlo (CMMC) method to efficiently perform sensitivity analysis of input parameters such as the failure rate of safety components. In the correlated sampling method, the original and the perturbed samples are assumed to trace an identical accident sequence, but the weight of the perturbed sample is adjusted to incorporate the variation of input data. The present method is applied to the sensitivity analysis of the safety evaluation of spent fuel pools. The result indicates that the sensitivity analysis for the CMMC coupling method can be efficiently carried out using the correlated sampling method.

    DOI: 10.1080/00223131.2023.2231464

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  18. Roadmap towards the introduction of new design fuels as well as the continuous improvement of safety evaluation (5)

    Murakami Kenta, Yamamoto Akio

    Journal of the Atomic Energy Society of Japan   Vol. 65 ( 10 ) page: 606 - 607   2023

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    Language:Japanese   Publishing type:Research paper (scientific journal)   Publisher:Atomic Energy Society of Japan  

    DOI: 10.3327/jaesjb.65.10_606

    CiNii Research

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

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

    Nuclear Science and Engineering   Vol. 196 ( 7 ) page: 769 - 791   2022

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    Language:English   Publishing type:Research paper (scientific journal)   Publisher: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

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

    Oshima Y., Endo T., Yamamoto A.

    Nuclear Science and Engineering   Vol. 196 ( 4 ) page: 379 - 394   2022

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    Language:English   Publishing type:Research paper (scientific journal)   Publisher: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

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

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

    Nuclear Science and Engineering   Vol. 196 ( 2 ) page: 133 - 143   2022

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    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

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

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

    Journal of Nuclear Science and Technology   Vol. 59 ( 9 ) page: 1117 - 1126   2022

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    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

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  23. Safety measures and risk management of nuclear power plants against external events (4)

    Murakami Kenta, Yamamoto Akio

    Journal of the Atomic Energy Society of Japan   Vol. 64 ( 5 ) page: 272 - 274   2022

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    DOI: 10.3327/jaesjb.64.5_272

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  24. Implementation of Resonance Upscattering Treatment in FRENDY Nuclear Data Processing System

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

    Nuclear Science and Engineering   Vol. 196 ( 11 ) page: 1267 - 1279   2022

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    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

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

    Katano R., Yamamoto A., Endo T.

    Nuclear Science and Engineering   Vol. 196 ( 10 ) page: 1194 - 1208   2022

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    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

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  26. 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   Vol. 64 ( 3 ) page: 176 - 176   2022

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    Language:Japanese   Publishing type:Research paper (scientific journal)   Publisher:一般社団法人 日本原子力学会  

    DOI: 10.3327/jaesjb.64.3_176

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  27. Toward construction of cross-organizational education system for nuclear engineering

    Kurosaki Ken, Kozaki Tamotsu, Nakashima Hiroshi, Obara Toru, Wakabayashi Genichiro, Pyeon Cheolho, Matsuyama Shigeo, Abe Hiroshi, Uno Masayoshi, Yamamoto Akio

    Journal of the Atomic Energy Society of Japan   Vol. 64 ( 9 ) page: 520 - 524   2022

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    DOI: 10.3327/jaesjb.64.9_520

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

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

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

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

    DOI: 10.13182/PHYSOR22-37693

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

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

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

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

    DOI: 10.13182/PHYSOR22-37287

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  30. Development of Nuclear Data Processing Code FRENDY Version 2

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

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

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

    DOI: 10.13182/PHYSOR22-37299

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  31. AREA RATIO METHOD USING DYNAMIC MODE DECOMPOSITION

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

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

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

    DOI: 10.13182/PHYSOR22-37184

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

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

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

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

    DOI: 10.13182/PHYSOR22-37251

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

    Katano R., Yamamoto A., Endo T.

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

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

    DOI: 10.13182/PHYSOR22-37557

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

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

    Scientific Reports   Vol. 11 ( 1 )   2021.12

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    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

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  35. Evolutionary simulated annealing for fuel loading optimization of VVER-1000 reactor International coauthorship

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

    Annals of Nuclear Energy   Vol. 151   2021.2

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    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

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  36. 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   Vol. 195 ( 1 ) page: 33 - 49   2021

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    DOI: 10.1080/00295639.2020.1781482

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  37. 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   Vol. 58 ( 2 ) page: 173 - 183   2021

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    DOI: 10.1080/00223131.2020.1814891

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

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

    Nuclear Science and Engineering   Vol. 195 ( 7 ) page: 694 - 716   2021

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    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

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  39. 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   Vol. 124 ( 1 ) page: 556 - 558   2021

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    Language:Japanese   Publishing type:Research paper (scientific journal)   Publisher:Transactions of the American Nuclear Society  

    DOI: 10.13182/T124-35116

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  40. 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   Vol. 124 ( 1 ) page: 184 - 187   2021

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    Language:Japanese   Publishing type:Research paper (scientific journal)   Publisher:Transactions of the American Nuclear Society  

    DOI: 10.13182/T124-35323

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  41. 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   Vol. 124 ( 1 ) page: 544 - 547   2021

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    Language:Japanese   Publishing type:Research paper (scientific journal)   Publisher:Transactions of the American Nuclear Society  

    DOI: 10.13182/T124-35126

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

    Yamamoto A., Endo T., Tada K.

    Journal of Nuclear Science and Technology   Vol. 58 ( 12 ) page: 1343 - 1350   2021

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    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

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

    Hayashi T., Endo T., Yamamoto A.

    Journal of Nuclear Science and Technology   Vol. 58 ( 9 ) page: 1008 - 1017   2021

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    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

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

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

    Journal of Nuclear Science and Technology   Vol. 58 ( 11 ) page: 1165 - 1183   2021

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    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

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

    Sawada K., Yamamoto A., Endo T., Sato C., Maeda K., Jang S.

    Journal of Nuclear Science and Technology   Vol. 58 ( 12 ) page: 1308 - 1317   2021

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    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

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  46. Application of Regionwise Even-Parity Discontinuity Factor to the Multigroup Analog Monte-Carlo Method

    Oshima Y., Endo T., Yamamoto A., Aizawa N.

    Transactions of the American Nuclear Society   Vol. 125 ( 1 ) page: 904 - 907   2021

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    DOI: 10.13182/T125-36788

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  47. Verification of Medium-Wise Multi-Group Cross Section Calculation Capability of FRENDY/MG

    Chiba G., Yamamoto A.

    Transactions of the American Nuclear Society   Vol. 125 ( 1 ) page: 934 - 937   2021

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    DOI: 10.13182/T125-36801

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  48. Dimension-reduced Nuclear Data Adjustment Method based on the Bayesian Monte-Carlo Method

    Fukui Y., Endo T., Yamamoto A.

    Transactions of the American Nuclear Society   Vol. 125 ( 1 ) page: 900 - 903   2021

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    DOI: 10.13182/T125-36787

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  49. THEORETICAL DERIVATION OF UNIQUE COMBINATION-NUMBER FOR HIGHER ORDER NEUTRON CORRELATION FACTORS BASED ON PÁL-BELL EQUATION

    Endo T., Yamamoto A.

    Proceedings of the International Conference on Mathematics and Computational Methods Applied to Nuclear Science and Engineering, M and C 2021     page: 1568 - 1576   2021

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    Language:English   Publishing type:Research paper (scientific journal)   Publisher:Proceedings of the International Conference on Mathematics and Computational Methods Applied to Nuclear Science and Engineering, M and C 2021  

    The Pál-Bell equation is a master equation to describe the probability generating function (PGF) of neutron counts in the 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. Thereby, the unique combination numbers for the higher-order neutron-correlation factors for a near-critical state can be theoretically clarified. This knowledge is useful to judge whether a target system is a near-critical state or not using only a histogram (or factorial moments) of neutron counts.

    DOI: 10.13182/M&C21-33627

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  50. DEVELOPMENT OF ESTIMATION METHOD FOR PROMPT NEUTRON DECAY CONSTANT USING DYNAMIC MODE DECOMPOSITION

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

    Proceedings of the International Conference on Mathematics and Computational Methods Applied to Nuclear Science and Engineering, M and C 2021     page: 1719 - 1728   2021

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    Language:English   Publishing type:Research paper (scientific journal)   Publisher:Proceedings of the International Conference on Mathematics and Computational Methods Applied to Nuclear Science and Engineering, M and C 2021  

    To estimate the prompt neutron decay constant α corresponding to the fundamental mode in a subcritical system, we try applying the Dynamic Mode Decomposition (DMD) to time-series data of neutron counts obtained by the pulsed-neutron source (PNS) method. In this study, we newly develop an estimation method of α with the statistical uncertainty by the DMD and the random sampling method. The applicability of the proposed method is demonstrated by analyzing a PNS experiment carried out at the Kyoto University Critical Assembly (KUCA).

    DOI: 10.13182/M&C21-33631

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  51. FRENDY/MG: A MULTI-GROUP CROSS SECTION GENERATION MODULE USING ACE POINTWISE CROSS SECTIONS

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

    Proceedings of the International Conference on Mathematics and Computational Methods Applied to Nuclear Science and Engineering, M and C 2021     page: 710 - 720   2021

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    Language:English   Publishing type:Research paper (scientific journal)   Publisher:Proceedings of the International Conference on Mathematics and Computational Methods Applied to Nuclear Science and Engineering, M and C 2021  

    A generation capability of neutron multi-group cross sections is being implemented to the FRENDY nuclear data processing code, as FRENDY/MG. FRENDY/MG generates neutron multi-group cross sections for deterministic core analysis codes considering an arbitrary energy group structure. Distinguished features of FRENDY/MG are 1) use of ACE pointwise cross sections as the source of nuclear data (no evaluated nuclear data file is directly used), 2) treatment of a compound material consisting of multiple nuclides to explicitly consider the resonance interference effect. Various verifications are being carried out through the comparison with the multi-group cross sections generated by NJOY.

    DOI: 10.13182/M&C21-33679

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  52. FUEL ASSEMBLY ANALYSES WITH RESONANCE CALCULATION USING ENERGY SPECTRUM EXPANSION METHOD

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

    Proceedings of the International Conference on Mathematics and Computational Methods Applied to Nuclear Science and Engineering, M and C 2021     page: 2219 - 2230   2021

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    Language:English   Publishing type:Research paper (scientific journal)   Publisher:Proceedings of the International Conference on Mathematics and Computational Methods Applied to Nuclear Science and Engineering, M and C 2021  

    Efficient numerical algorithms for a Resonance calculation using energy Spectrum Expansion (RSE) method were proposed for a large and complicated geometry such as a fuel assembly. The RSE method can treat complicated heterogeneous geometry considering resonance interference effect among different regions. Calculation procedure of the RSE method is composed of (1)generation of ultra-fine group spectra from ultra-fine group calculations in homogeneous geometry, (2)expansion of the spectra by orthogonal basis on energy based on singular value decomposition, (3)transport calculation for expansion coefficients and (4)reconstruction of ultra-fine group spectra in target heterogeneous regions by the expansion coefficients and the orthogonal basis. In this study, the efficient numerical algorithms for the RSE method were developed and applied. The algorithms are composed of (i)scattering source calculation by the combination of conventional slowing down and moment-to-moment transfer calculations, and (ii)matrix exponential generation in transport calculation of the RSE method with method of characteristics by diagonalization using eigenvalue decomposition. Through numerical verification, it was confirmed that effective multi-group cross sections by the RSE method with the new algorithms are well agreed with those by direct ultra-fine group calculation in the fuel assembly geometry. The RSE method with the new algorithms is applicable to the fuel assembly analysis.

    DOI: 10.13182/M&C21-33621

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  53. IMPLEMENTATION OF A RESONANCE CALCULATION USING ENERGY SPECTRUM EXPANSION METHOD INTO HETEROGENEOUS TRANSPORT CALCULATION CODE GALAXY-Z

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

    Proceedings of the International Conference on Mathematics and Computational Methods Applied to Nuclear Science and Engineering, M and C 2021     page: 1982 - 1991   2021

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    Language:English   Publishing type:Research paper (scientific journal)   Publisher:Proceedings of the International Conference on Mathematics and Computational Methods Applied to Nuclear Science and Engineering, M and C 2021  

    A resonance calculation using the energy spectrum expansion (RSE) method has been developed in order to efficiently treat complicated heterogeneous geometry and resonance interference effect among both nuclides and regions. The three-dimensional (3D) heterogeneous transport calculation code GALAXY-Z developed by Mitsubishi Heavy Industries Ltd. uses the RSE method as resonance calculation method. In this study, resonane calculations of the GALAXY-Z code using the RSE method were performed and showed good agreements of nuclear characteristics with the ultra-fine-group method in wide range of moderator density for actual PWR fuel specifications. The GALAXY-Z code using the RSE method is applicable with high accuracy to generation of nuclear constants in normal operations and accident conditions including low moderator density conditions and can siginificantly reduce computational memory in comparison with the ultra-fine-group method.

    DOI: 10.13182/M&C21-33788

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  54. S2 Consistent Analytic CMFD Acceleration for Method of Characteristics

    Oshima Y., Yamamoto A., Endo T.

    Proceedings of the International Conference on Mathematics and Computational Methods Applied to Nuclear Science and Engineering, M and C 2021     page: 1840 - 1849   2021

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    Language:English   Publishing type:Research paper (scientific journal)   Publisher:Proceedings of the International Conference on Mathematics and Computational Methods Applied to Nuclear Science and Engineering, M and C 2021  

    A coarse mesh finite-difference (CMFD) acceleration consistent with the S2 discrete ordinate method using the step-characteristics is proposed. The analytically derived finite-difference formulation for diffusion theory (ACMFD), which is theoretically consistent with the discrete ordinate method in one-dimensional slab geometry with the S2 Gauss-Legendre quadrature, is used for the acceleration of a transport calculation. Our numerical calculations in simple geometry revealed that the consistent treatment of scattering source distribution in the ACMFD acceleration has a significant impact on the convergence property. Further, the linearized Fourier analysis of ACMFD acceleration shows the same convergence property as numerical calculations. These numerical and theoretical analyses show that the S2- consistent ACMFD acceleration is unconditionally stable without any correction to the diffusion coefficient.

    DOI: 10.13182/M&C21-33737

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

    Endo Tomohiro, Yamamoto Akio

    NUCLEAR SCIENCE AND ENGINEERING   Vol. 194 ( 11 ) page: 1089 - 1104   2020.11

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    DOI: 10.1080/00295639.2020.1720499

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  56. Implementation of the unscented transformation with low rank approximation in uncertainty analysis during large-break loss of coolant accident International coauthorship

    Foad Basma, Yamamoto Akio, Endo Tomohiro

    ANNALS OF NUCLEAR ENERGY   Vol. 146   2020.10

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    DOI: 10.1016/j.anucene.2020.107614

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  57. Uncertainty and regression analysis of the MSLB accident in PWR based on unscented transformation and low rank approximation International coauthorship

    Foad Basma, Yamamoto Akio, Endo Tomohiro

    ANNALS OF NUCLEAR ENERGY   Vol. 143   2020.8

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    DOI: 10.1016/j.anucene.2020.107493

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  58. Applicability of a reduced order model for a safety analysis code to statistical safety analysis Reviewed

    Matsushita Masaki, Endo Tomohiro, Yamamoto Akio

    JOURNAL OF NUCLEAR SCIENCE AND TECHNOLOGY     page: 1-12   2020.7

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    DOI: 10.1080/00223131.2020.1783382

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  59. Efficient uncertainty quantification for PWR during LOCA using unscented transform with singular value decomposition International coauthorship

    Foad Basma, Yamamoto Akio, Endo Tomohiro

    ANNALS OF NUCLEAR ENERGY   Vol. 141   2020.6

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    DOI: 10.1016/j.anucene.2020.107341

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  60. 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   Vol. 194 ( 6 ) page: 477 - 491   2020.6

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    DOI: 10.1080/00295639.2020.1722512

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  61. Application of the multigrid amplitude function method for time-dependent MOC based on the linear source approximation Reviewed

    Tsujita Kosuke, Endo Tomohiro, Yamamoto Akio

    JOURNAL OF NUCLEAR SCIENCE AND TECHNOLOGY     2020.1

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    DOI: 10.1080/00223131.2019.1709993

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  62. Subcriticality measurement using time-domain decomposition-based integral method for simultaneous reactivity and source changes Reviewed

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

    JOURNAL OF NUCLEAR SCIENCE AND TECHNOLOGY     2020.1

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    DOI: 10.1080/00223131.2019.1706658

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  63. A kinetic model for the direct response matrix method

    Mitsuyasu Takeshi, Yamamoto Akio

    JOURNAL OF NUCLEAR SCIENCE AND TECHNOLOGY   Vol. 57 ( 1 ) page: 90 - 99   2020.1

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    DOI: 10.1080/00223131.2019.1659874

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  64. 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   Vol. 123 ( 1 ) page: 1349 - 1353   2020

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    Language:Japanese   Publishing type:Research paper (scientific journal)   Publisher:Transactions of the American Nuclear Society  

    DOI: 10.13182/T123-33370

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  65. 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   Vol. 123 ( 1 ) page: 841 - 844   2020

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    DOI: 10.13182/T123-33367

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  66. Recent Research Activities on Nuclear Reactor Physics Using the Computational Science

    YAMAMOTO Akio

    The Proceedings of Mechanical Engineering Congress, Japan   Vol. 2020 ( 0 ) page: K08101   2020

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    <p>Outline of the nuclear reactor core analysis and computational methods used in this analysis are described. Various techniques and approximations are used to perform core analysis with practical computational resources. Recent progress on computer hardware and numerical algorithms enabled us to use more precise and accurate analysis methods.</p>

    DOI: 10.1299/jsmemecj.2020.k08101

    CiNii Research

  67. 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   Vol. 2020-March   page: 152 - 161   2020

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    Language:Japanese   Publishing type:Research paper (scientific journal)   Publisher: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

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  68. 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   Vol. 2020-March   page: 1736 - 1744   2020

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    Language:Japanese   Publishing type:Research paper (scientific journal)   Publisher: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

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  69. 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   Vol. 2020-March   page: 2690 - 2697   2020

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    Language:Japanese   Publishing type:Research paper (scientific journal)   Publisher: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

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  70. 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   Vol. 122   page: 714 - 717   2020

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    DOI: 10.13182/T122-32047

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  71. 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   Vol. 122   page: 458 - 461   2020

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    DOI: 10.13182/T122-32039

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  72. Loading pattern optimization for a PWR using Multi-Swarm Moth Flame Optimization Method with Predator Reviewed

    Ishiguro Satomi, Endo Tomohiro, Yamamoto Akio

    JOURNAL OF NUCLEAR SCIENCE AND TECHNOLOGY     2019.12

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    DOI: 10.1080/00223131.2019.1700844

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  73. Development of a reduced order model for severe accident analysis codes by singular value decomposition aiming probabilistic safety margin analysis Reviewed

    Matsushita Masaki, Endo Tomohiro, Yamamoto Akio

    JOURNAL OF NUCLEAR SCIENCE AND TECHNOLOGY     2019.12

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    DOI: 10.1080/00223131.2019.1699190

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  74. Track 1: Severe Accident Phenomena, Fukushima Accident Analysis

    Journal of the Society of Mechanical Engineers   Vol. 122 ( 1211 ) page: 10 - 12   2019.10

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    DOI: 10.1299/jsmemag.122.1211_10

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  75. APPLICATION OF THE FOREST SHIELDING FACTOR TO THE MAXIMUM-LIKELIHOOD EXPECTATION MAXIMIZATION METHOD FOR AIRBORNE RADIATION MONITORING Reviewed

    Sasaki M., Sanada Y., Yamamoto A.

    RADIATION PROTECTION DOSIMETRY   Vol. 184 ( 3-4 ) page: 400-404   2019.10

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    DOI: 10.1093/rpd/ncz095

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

    Yamamoto Akio, Endo Tomohiro, Giho Akinori

    JOURNAL OF NUCLEAR SCIENCE AND TECHNOLOGY   Vol. 56 ( 8 ) page: 716 - 723   2019.8

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    DOI: 10.1080/00223131.2019.1618405

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  77. 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   Vol. 56 ( 6 ) page: 471 - 478   2019.6

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    DOI: 10.1080/00223131.2019.1598509

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  78. 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   Vol. 2019-May   2019.5

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    Language:Japanese   Publishing type:Research paper (scientific journal)   Publisher: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.

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  79. 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   Vol. 56 ( 4 ) page: 322 - 336   2019.4

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    DOI: 10.1080/00223131.2019.1580625

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

    Yamamoto Akio, Giho Akinori, Endo Tomohiro

    NUCLEAR SCIENCE AND ENGINEERING   Vol. 193 ( 3 ) page: 253 - 268   2019.3

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    DOI: 10.1080/00295639.2018.1516961

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

    Endo Tomohiro, Yamamoto Akio

    ANNALS OF NUCLEAR ENERGY   Vol. 124   page: 606 - 615   2019.2

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    DOI: 10.1016/j.anucene.2018.10.032

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

    Sugaya Shinji, Endo Tomohiro, Yamamoto Akio

    ANNALS OF NUCLEAR ENERGY   Vol. 124   page: 49 - 57   2019.2

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    DOI: 10.1016/j.anucene.2018.09.022

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  83. A New Interpretation of Discontinuity Factor

    Yamamoto Akio, Endo Tomohiro

    NUCLEAR SCIENCE AND ENGINEERING   Vol. 193 ( 9 ) page: 991 - 997   2019

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    DOI: 10.1080/00295639.2019.1579514

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  84. 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)   Vol. 2019.27 ( 0 ) page: 1022   2019

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    Language:English   Publishing type:Research paper (scientific journal)   Publisher: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

  85. Subcriticality - from basics to applications (1)

    Endo Tomohiro, Tsujimoto Kazufumi, Yamamoto Akio

    Journal of the Atomic Energy Society of Japan   Vol. 61 ( 10 ) page: 734 - 738   2019

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    Language:Japanese   Publishing type:Research paper (scientific journal)   Publisher:Atomic Energy Society of Japan  

    DOI: 10.3327/jaesjb.61.10_734

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  86. Mission of Reactor Physics

    Chiba Go, Yamamoto Akio

    Journal of the Atomic Energy Society of Japan   Vol. 61 ( 4 ) page: 254 - 256   2019

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    Language:Japanese   Publishing type:Research paper (scientific journal)   Publisher:Atomic Energy Society of Japan  

    DOI: 10.3327/jaesjb.61.4_254

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  87. 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   Vol. 120   page: 851 - 854   2019

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  88. 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     page: 1864 - 1873   2019

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    Language:Japanese   Publishing type:Research paper (scientific journal)   Publisher: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.

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  89. 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   Vol. 121   page: 979 - 982   2019

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    Language:Japanese   Publishing type:Research paper (scientific journal)   Publisher:Transactions of the American Nuclear Society  

    DOI: 10.13182/T30744

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  90. 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     page: 1493 - 1502   2019

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    Language:Japanese   Publishing type:Research paper (scientific journal)   Publisher: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.

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  91. 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   Vol. 121   page: 1309 - 1312   2019

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    DOI: 10.13182/T30783

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  92. 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   Vol. 121   page: 1316 - 1320   2019

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    DOI: 10.13182/T30993

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  93. 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     page: 1874 - 1885   2019

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    Language:Japanese   Publishing type:Research paper (scientific journal)   Publisher: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.

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  94. 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     page: 1081 - 1092   2019

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    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.

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  95. 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     page: 1534 - 1543   2019

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    Language:Japanese   Publishing type:Research paper (scientific journal)   Publisher: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.

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  96. 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   Vol. 55 ( 12 ) page: 1434 - 1458   2018.12

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    DOI: 10.1080/00223131.2018.1513875

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  97. 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   Vol. 192 ( 3 ) page: 240 - 253   2018.12

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    DOI: 10.1080/00295639.2018.1501978

    Web of Science

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  98. 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   Vol. 122   page: 146 - 154   2018.12

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    DOI: 10.1016/j.anucene.2018.08.041

    Web of Science

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

    Endo Tomohiro, Yamamoto Akio

    JOURNAL OF NUCLEAR SCIENCE AND TECHNOLOGY   Vol. 55 ( 11 ) page: 1245 - 1254   2018.11

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    DOI: 10.1080/00223131.2018.1491902

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  100. Reduction of Cross Section Table Size for Core Analysis Using Dimensionality Reduction Technique Reviewed

    M. Yamamoto, T. Endo, A. Yamamoto

    Trans. Am. Nucl. Soc.   Vol. 119   page: 1226-1228   2018.11

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  101. A Simple Treatment of Bowed Assembly Gap Through Correction of Cross Section Reviewed

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

    Trans. Am. Nucl. Soc.   Vol. 119   page: 1199-1202   2018.11

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  102. Transport Consistent Diffusion Coefficient for CMFD Acceleration Reviewed

    A. Yamamoto, T. Endo

    Trans. Am. Nucl. Soc.   Vol. 119   page: 1179-1181   2018.11

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  103. Estimation of Subcriticality in Dollar Units Based on Integral Method for Arbitrary State-Change in Subcritical System Reviewed

    A. Nonaka, T. Endo, A. Yamamoto

    Trans. Am. Nucl. Soc.   Vol. 119   page: 1112-1115   2018.11

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  104. Surrogate Model of Severe Accident Analysis Code for SBO Aiming Probabilistic Safety Margin Analysis Reviewed

    M. Matsushita, T. Endo, A. Yamamoto

    Trans. Am. Nucl. Soc.   Vol. 119   page: 900-903   2018.11

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  105. 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   Vol. 55 ( 10 ) page: 1099 - 1109   2018.10

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    DOI: 10.1080/00223131.2018.1479988

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  106. Estimation of Subcriticality using Particle Filter Method Reviewed

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

    Trans. Am. Nucl. Soc.   Vol. 118   page: 851-854   2018.6

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  107. Quantification of Modeling Approximation Error of Pin-Cell Calculation Using Kriging and Principal Component Analysis Reviewed

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

    Trans. Am. Nucl. Soc.   Vol. 118   page: 875-878   2018.6

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  108. Inverse Estimation Methods of Unknown Radioactive Source for Fuel Debris Search Reviewed

    S. Sugaya, T. Endo, A. Yamamoto

    Proc. PHYSOR201     page: [USB-DRIVE]   2018.4

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  109. Estimation of Subcriticality in Dollar Units using Integral Method for Subcritical System Reviewed

    A. Nonaka, T. Endo, A. Yamamoto

    Proc. PHYSOR2018     page: [USB-DRIVE]   2018.4

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  110. Cache Efficient Flux Region Assignment for the Method of Characteristics Reviewed

    A. Yamamoto, A. Giho, T. Endo

    Proc. PHYSOR2018     page: [USB-DRIVE]   2018.4

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  111. Estimation of Region-Wise Even-Parity Discontinuity Factor for MOC Through Iterative Procedure Reviewed

    A. Yamamoto, A. Giho, T. Endo

    Proc. PHYSOR2018     page: [USB-DRIVE]   2018.4

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  112. 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     page: [USB-DRIVE]   2018.4

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  113. Sensitivity Coefficient Analysis of Omega-Eigenvalue based on First-Order Perturbation Theory Reviewed

        page: [USB-DRIVE]   2018.4

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

    Yokoyama Kenji, Yamamoto Akio, Kitada Takanori

    JOURNAL OF NUCLEAR SCIENCE AND TECHNOLOGY   Vol. 55 ( 3 ) page: 319 - 334   2018.3

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    DOI: 10.1080/00223131.2017.1397563

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  115. 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   Vol. 55 ( 1 ) page: 41 - 65   2018.1

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    DOI: 10.1080/00223131.2017.1384704

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  116. Utilization of Region-wise Even-Parity Discontinuity Factor to Reduce Discretization Error of MOC Reviewed

    A. Yamamoto, A. Giho, T. Endo

    Nuclear Science and Engineering   Vol. 193   page: 253-268   2018

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  117. 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   Vol. 192   page: 243-250   2018

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  118. Estimation of Sensitivity Coefficient based on Lasso-type Penalized Linear Regression Reviewed

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

    JOURNAL OF NUCLEAR SCIENCE AND TECHNOLOGY   Vol. 55   page: 1099-1109   2018

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  119. Symposium Report on ASRAM2017

    Yamaguchi Akira, Yamamoto Akio, Narumiya Yoshiyuki

    Journal of the Atomic Energy Society of Japan   Vol. 60 ( 6 ) page: 362 - 363   2018

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    DOI: 10.3327/jaesjb.60.6_362

    CiNii Research

  120. Research and development roadmap for reactor physics 2017:future of reactor physics projected by next generations

    Yamamoto Akio, Chiba Go, Kirimura Kazuki, Miki Yosuke, Yokoyama Kenji

    Journal of the Atomic Energy Society of Japan   Vol. 60 ( 4 ) page: 241-245   2018

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    DOI: 10.3327/jaesjb.60.4_241

  121. 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   Vol. 119   page: 900 - 903   2018

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  122. 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   Vol. Part F168384-5   page: 3042 - 3053   2018

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    Language:Japanese   Publishing type:Research paper (scientific journal)   Publisher: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.

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  123. 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   Vol. Part F168384-3   page: 1793 - 1804   2018

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    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.

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  124. 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   Vol. Part F168384-3   page: 1892 - 1903   2018

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    Language:Japanese   Publishing type:Research paper (scientific journal)   Publisher: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.

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  125. 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   Vol. 119   page: 1112 - 1115   2018

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  126. 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   Vol. Part F168384-5   page: 3271 - 3282   2018

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    Language:Japanese   Publishing type:Research paper (scientific journal)   Publisher: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.

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  127. Estimation of subcriticality using particle filter method

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

    Transactions of the American Nuclear Society   Vol. 118   page: 851 - 854   2018

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  128. Estimation of subcriticality using particle filter method

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

    AISTech - Iron and Steel Technology Conference Proceedings   Vol. 2018-May   page: 851 - 854   2018

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  129. 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   Vol. Part F168384-4   page: 2632 - 2643   2018

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    Language:Japanese   Publishing type:Research paper (scientific journal)   Publisher: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.

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  130. 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   Vol. 118   page: 875 - 878   2018

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  131. 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   Vol. 2018-May   page: 875 - 878   2018

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  132. 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   Vol. 119   page: 1226 - 1228   2018

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  133. 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   Vol. Part F168384-2   page: 1240 - 1253   2018

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    Language:Japanese   Publishing type:Research paper (scientific journal)   Publisher: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.

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  134. 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   Vol. Part F168384-2   page: 788 - 799   2018

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    Language:Japanese   Publishing type:Research paper (scientific journal)   Publisher: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.

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  135. Estimation of Modeling Approximation Error of Core Analysis Using the Surrogate Model with Kriging Reviewed

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

    rans. Am. Nucl. Soc.   Vol. 117   page: 1269-1272   2017.11

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  136. Application of the GENESIS Code to the Kobayashi 3D Benchmark Problem Reviewed

    Akio Yamamoto, Akinori Giho, Tomohiro Endo

    Trans. Am. Nucl. Soc.   Vol. 117   page: 1403-1406   2017.11

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  137. 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   Vol. 49 ( 6 ) page: 1143 - 1156   2017.9

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    DOI: 10.1016/j.net.2017.06.016

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  138. Uncertainty Quantification of Activation Due to Cross Section Data in Neutron Shielding Calculation Reviewed

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

    Proc. 2017 International. Cogress on Advances in Nuclear Power Plants     2017.4

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    DOI: CD-ROM

  139. Application of Bias Factor Method using Random Sampling Technique for Critical Eigenvalue Prediction of BWR Reviewed

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

    Proc. 2017 International Cogress on Advances in Nuclear Power Plants     2017.4

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  140. Development of Dynamic Probabilistic Risk Assessment Model for PWR using Simplified Plant Simulation Method Reviewed

    Shohei Otsuki,Tomohiro Endo,Akio Yamamoto

    Proc. 2017 International Cogress on Advances in Nuclear Power Plants     2017.4

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  141. Development of GENESIS, a Three-dimensional Heterogeneous Transport Code based on the LEAF Method Reviewed

    Akio Yamamoto,Akinori Giho,Tomohiro Endo

    Proc. Int. Conf. on Math. and Comp. Methods Applied to Nucl. Sci. & Eng. (M&C2017)     2017.4

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    DOI: USB-DRIVE

  142. Theoretical Discussion of Statistical Error of Variance-to-Mean Ratio Reviewed

    Tomohiro Endo,Akio Yamamoto

    Proc. Int. Conf. on Math. and Comp. Methods Applied to Nucl. Sci. & Eng. (M&C2017)     2017.4

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    DOI: USB-DRIVE

  143. 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)     2017.4

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    DOI: USB-DRIVE

  144. Uncertainty Quantification of Bell Factor for Sjöstrand Method due to Cross-section Data for MA core Reviewed

    Toshiki Kimura,Tomohiro Endo,Akio Yamamoto

    Proc. Int. Conf. on Math. and Comp. Methods Applied to Nucl. Sci. & Eng. (M&C2017)     2017.4

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    DOI: USB-DRIVE

  145. 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   Vol. 54 ( 4 ) page: 459 - 471   2017.4

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    DOI: 10.1080/00223131.2017.1286271

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  146. 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   Vol. 102   page: 77 - 84   2017.4

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    Language:English   Publishing type:Research paper (scientific journal)  

    DOI: 10.1016/j.anucene.2016.11.045

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  147. Automated generation of burnup chain for reactor analysis applications

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

    KERNTECHNIK   Vol. 82 ( 2 ) page: 196 - 205   2017.4

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    Language:English   Publishing type:Research paper (scientific journal)  

    DOI: 10.3139/124.110671

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  148. 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   Vol. 186 ( 1 ) page: 1 - 22   2017.4

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    Language:English   Publishing type:Research paper (scientific journal)  

    DOI: 10.1080/00295639.2016.1273002

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  149. Estimation of Modeling Approximation Errors using Data Assimilation with the Minimum Variance Approach Reviewed

    A. Yamamoto,K. Kinoshita,T. Endo

    J. Nucl. Sci. Technol.   Vol. 54   page: 459-471   2017.2

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  150. 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   Vol. 54 ( 1 ) page: 30 - 38   2017.1

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    DOI: 10.1080/00223131.2016.1202793

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  151. GENESIS - A Three-dimensional Heterogeneous Transport Solver based on the Legendre Polynomial Expansion of Angular Flux Method Reviewed

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

    Nucl. Sci. Eng.   Vol. 186   page: 1-22   2017

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  152. A Coupling Model for the Two-stage Core Calculation Method with Subchannel Analysis for Boiling Water Reactors Reviewed

    T. Mitsuyasu, M. Aoyama, A. Yamamoto

    Ann. Nucl. Energy   Vol. 102   page: 77-84   2017

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  153. Estimation of Sensitivity Coefficients of Core Characteristics based on Reduced-order Modeling using Sensitivity Matrix of Assembly Characteristics Reviewed

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

    J. Nucl. Sci. Technol.   Vol. 54   page: 637-647   2017

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  154. Application of the GENESIS Code to the Kobayashi 3D Benchmark Problem Reviewed

    Akio Yamamoto,Akinori Giho,Tomohiro Endo

    Trans. Am. Nucl. Soc.   Vol. 117   page: 1403-1406   2017

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  155. Estimation of Modeling Approximation Error of Core Analysis Using the Surrogate Model with Kriging Reviewed

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

    Trans. Am. Nucl. Soc.   Vol. 117   page: 1269-1272   2017

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  156. Application of the Bias Factor Method Using the Random Sampling Technique for Prediction Accuracy Improvement of Neutronics Parameters of BWR Reviewed

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

    Trans. Am. Nucl. Soc.   Vol. 117   page: 804-807   2017

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

    A. Yamamoto,A. Giho,T. Endo

    Nucl. Eng. Technol.   Vol. 49   page: 1143-1156   2017

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  158. Automated Generation of Burnup Chain for Reactor Analysis Applications Reviewed

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

    Kerntechnik   Vol. 82   page: 196-205   2017

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  159. 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   Vol. 54 ( 6 ) page: 637-647   2017

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    DOI: 10.1080/00223131.2017.1299052

    Web of Science

  160. 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

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    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

  161. 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

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    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

  162. 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

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    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.

    Scopus

  163. 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   Vol. 117   page: 1269 - 1272   2017

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    Scopus

  164. 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

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    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

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

    Yamamoto A., Giho A., Endo T.

    Transactions of the American Nuclear Society   Vol. 117   page: 1403 - 1406   2017

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    Scopus

  166. GENESIS - a Transport Solver in Three-Dimensional Heterogeneous Geometry based on the Leaf Method Reviewed

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

    Proc. PHYSOR2016     2016.4

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  167. Statistical Error Estimation of the Feynman-α Method using the Bootstrap Method Reviewed

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

    J. Nucl. Sci. Technol.   Vol. 53   page: 1447-1453   2016

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  168. Discontinuity Factors for Simplified P3 Theory Reviewed

    A. Yamamoto,T. Sakamoto,T. Endo

    Nucl. Sci. Eng.   Vol. 183   page: 39-51   2016

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  169. Bias Factor Method Using Random Sampling Technique Reviewed

    T. Endo,A. Yamamoto,T. Watanabe

    J. Nucl. Sci. Technol.   Vol. 53   page: 1491-1501   2016

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  170. Cross Section Adjustment Methods based on Minimum Variance Unbiased Estimation Reviewed

    K. Yokoyama,A. Yamamoto

    J. Nucl. Sci. Technol.   Vol. 53   page: 1622-1638   2016

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

    T. Okubo,A. Yamamoto,E. Endo

    J. Nucl. Sci. Technol.   Vol. 53   page: 1-9   2016

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  172. Reduction of MOC Discretization Errors Through a Minimization of Source Ratio Variances Reviewed

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

    J. Nucl. Sci. Technol.   Vol. 53   page: 1858-1869   2016

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  173. A CMFD Acceleration Method for SP3 Advanced Nodal Method Reviewed

    A. Yamamoto,T. Sakamoto,T. Endo

    Nucl. Sci. Eng.   Vol. 184   page: 168-173   2016

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  174. Uncertainty Quantification of Spatial Correction Factor for Sjöstrand Method due to Cross-Section Data Reviewed

    T. Kimura,T. Endo,A. Yamamoto

    Trans. Am. Nucl. Soc.   Vol. 115   page: 1081-1084   2016

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  175. Comparison of Fuel Loading Pattern Optimization Results using Exhaustive Search for Fresh Fuels and Local Search for Burned Fuels Reviewed

    S. Ishiguro,T. Endo,A. Yamamoto

    Trans. Am. Nucl. Soc.   Vol. 115   page: 1265-1267   2016

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  176. Comparison of the Numerical Stability between CMFD and GCMR with Stabilization Techniques Reviewed

    Akinori Giho,A. Yamamoto

    Trans. Am. Nucl. Soc.   Vol. 115   page: 1245-1248   2016

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

    A. Yamamoto,Akinori Giho,T. Endo

    Trans. Am. Nucl. Soc.   Vol. 115   page: 1241-1244   2016

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  178. Prediction on Underestimation of Statistical Uncertainty in Monte Carlo Eigenvalue Calculation for Two-Dimensional Heterogeneous Color Set Geometry Reviewed

    K. Hayashi,T. Endo,A. Yamamoto

    Trans. Am. Nucl. Soc.   Vol. 115   page: 1213-1216   2016

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  179. Uncertainty Quantification of Activation Due to Cross Section Data in Neutron Shielding Calculation Reviewed

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

    Trans. Am. Nucl. Soc.   Vol. 115   page: 1085-1087   2016

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  180. Development of a Simplified Estimation Method on Severe Accident Progression in PWR for Education Reviewed

    S. Otsuki,T. Endo,A. Yamamoto

    Trans. Am. Nucl. Soc.   Vol. 113   page: 855-858   2015.11

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  181. Application of Data Assimilation based on Bayesian Theory to Subcriticality Measurements using Area Ratio Method Reviewed

    K. Maeno,T. Endo,A. Yamamoto

    Trans. Am. Nucl. Soc.   Vol. 113   page: 1282-1286   2015.11

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  182. Theoretical Expression of Area Ratio Method Using Detected-Neutron Multiplication Factor Reviewed

    T. Endo,A. Yamamoto

    Trans. Am. Nucl. Soc.   Vol. 113   page: 1208-1211   2015.11

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  183. Application of Partially-Converged Solution of Assembly Calculation for Core Sensitivity Analysis based on Reduced Order Modeling Reviewed

    R. Katano,T. Endo,A. Yamamoto

    Trans. Am. Nucl. Soc.   Vol. 113   page: 1161-1164   2015.11

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  184. Underestimation of Statistical Uncertainty of Monte Carlo Method with Non-Analog of Fission Source Sampling Reviewed

    H. Hayashi,T. Endo,A. Yamamoto

    Trans. Am. Nucl. Soc.   Vol. 113   page: 1153-1157   2015.11

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  185. Uncertainty Estimation of Analysis Model using the Data Assimilation Method Reviewed

    K. Kinoshita,T. Endo,A. Yamamoto

    Trans. Am. Nucl. Soc.   Vol. 113   page: 1141-1143   2015.11

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  186. Discontinuity Factors for Simplified P3 Theory Reviewed

    A.Yamamoto,T. Sakamoto,T. Endo

    Proc. Reactor Physics Asia 2015 (RPHA15),Sep. 17-18, Jeju, Korea     2015.9

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  187. Reactor Physics Activities in Nagoya University Reviewed

    A. Yamamoto,T. Endo

    Proc. Reactor Physics Asia 2015 (RPHA15), Sep. 17-18, Jeju, Korea     2015.9

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  188. Development of New Statistical Geometry Model using the Delta-tracking Method Reviewed

    T. Koide,T. Endo,A. Yamamoto

    Proc. Reactor Physics Asia 2015 (RPHA15), Sep. 17-18, Jeju, Korea     2015.9

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  189. Efficient Execution of Monte Carlo Simulation Based on Pseudo-Scattering using GPU Reviewed

    T. Okubo,T. Endo,A. Yamamoto

    Trans. Am. Nucl. Soc.   Vol. 112   page: 652-656   2015.6

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  190. Angular Dependent Transmission Probability Method for Fast Reactor Core Transport Analysis Reviewed

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

    Trans. Am. Nucl. Soc.   Vol. 112   page: 736-738   2015.6

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  191. Development of MHI FBR Nuclear Design Code System CALAXY-H/ENSEMBLE-TRIZ Reviewed

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

    Trans. Am. Nucl. Soc.   Vol. 112   page: 733-735   2015.6

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  192. Development of Core Sensitivity Analysis Based on Reduced-Order Modeling using Assembly Calculations Reviewed

    R. Katano,A. Yamamoto,T. Endo

    Trans. Am. Nucl. Soc.   Vol. 112   page: 715-718   2015.6

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

    T. Fujita,T. Endo,A. Yamamoto

    J. Nucl. Sci. Technol.   Vol. 52   page: 355-370   2015.3

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  194. Uncertainty Quantification of LWR Core Characteristics using Random Sampling Method Reviewed

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

    Nucl. Sci. Eng.   Vol. 181   page: 160-174   2015

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  195. Integration of equivalence theory and ultra-fine-group slowing-down calculation for resonance self-shielding treatment in lattice physics code GALAXY Reviewed

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

    J. Nucl. Sci. Technol.   Vol. 52   page: 842-869   2015

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  196. Confidence interval estimation by bootstrap method for uncertainty quantification using random sampling method Reviewed

    T. Endo,T. Watanabe,A. Yamamoto

    J. Nucl. Sci. Technol.   Vol. 52   page: 993-999   2015

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  197. Statistical Error Estimation Using Bootstrap Method for the Feynman-alpha Method Reviewed

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

    Trans. Am. Nucl. Soc.   Vol. 111   page: 1204-1207   2014.11

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  198. Development of Overall Safety Evaluation Method for Operating Nuclear Power Plant Considering Aging Effects -Concept and Framework- Reviewed

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

    Proc. ICMST-Kobe 2014     2014.11

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  199. Development of Overall Safety Evaluation Method for Operating Nuclear Power Plant Considering Aging Effects - New Risk Indicators for Maintenance Procedure- Reviewed

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

    Proc. ICMST-Kobe 2014     2014.11

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  200. Comparison of Spatial Source Expansion Methods in the Three Dimensional Transport Method LEAF Reviewed

    Y. Kato,T. Endo,A. Yamamoto

    Trans. Am. Nucl. Soc.   Vol. 111   page: 1405-1408   2014.11

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  201. Estimation of Sensitivity Coefficient using Random Sampling and L1-norm Minimization Reviewed

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

    Trans. Am. Nucl. Soc.   Vol. 111   page: 1391-1394   2014.11

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  202. Confidence Interval Estimation by Bootstrap Method for Uncertainty Quantification using Random Sampling Reviewed

    T. Endo,T. Watanabe,A. Yamamoto

    Proc. PHYSOR2014     2014.9

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  203. Investigation on Subcriticality Measurement using Inherent Neutron Source in Nuclear Fuel Reviewed

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

    Proc. PHYSOR2014     2014.9

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  204. Impact of Nearest Neighbor Distribution of Fuel Particle on Neutronics Characteristics in Statistical Geometry Model Reviewed

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

    Proc. PHYSOR2014     2014.9

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  205. Theoretical Prediction on Underestimation of Statistical Uncertainty for Fission Rate Tally in Monte Carlo Calculation Reviewed

    T. Endo.A. Yamamoto,K. Sakata

    Proc. PHYSOR2014     2014.9

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  206. Development of Legendre expansion of Angular Flux Method For 3D MOC Calculation Reviewed

    Y. Kato,A. Yamamoto,T. Endo

    Proc. PHYSOR2014     2014.9

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  207. Improvement of a Convergence Technique for MOC Calculation with Large Negative Self-Scattering Cross Section Reviewed

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

    Proc. PHYSOR2014     2014.9

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  208. Generation of Simplified Burnup Chain using Contribution Matrix of Nuclide Production Reviewed

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

    Proc. PHYSOR2014     2014.9

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  209. Uncertainty Quantification of Neutronics Characteristics using Latin Hypercube Sampling Method Reviewed

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

    Proc. PHYSOR2014     2014.9

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  210. Applicability of the Cross Section Adjustment Method based on Random Sampling Technique For Burnup Reviewed

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

    Proc. PHYSOR2014     2014.9

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  211. Uncertainty Quantification of BWR Core Characteristics using Latin Hypercube Sampling Method Reviewed

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

    Proc. PHYSOR2014     2014.9

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  212. Prediction on Underestimation of Variance for Fission Rate Distribution in Monte-Carlo Calculation Reviewed

    A. Yamamoto,K. Sakata,T. Endo

    Trans. Am. Nucl. Soc.   Vol. 110   page: 515-518   2014.6

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  213. An Improved CMFD Acceleration for SP3 Advanced Nodal Method Reviewed

    T. Sakamoto,A. Yamamoto,T. Endo

    Trans. Am. Nucl. Soc.   Vol. 110   page: 535-537   2014.6

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  214. Applicability of angular flux discontinuity factor preserving region-wise leakage for integro-differential transport equation Reviewed

    T. Sakamoto,T. Endo,A. Yamamoto

    J. Nucl. Sci. Technol.   Vol. 51   page: 1264-1273   2014.6

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  215. Applicability of angular flux discontinuity factor preserving region-wise leakage for integro-differential transport equation Reviewed

    T. Sakamoto,T. Endo,A. Yamamoto

    J. Nucl. Sci. Technol.   Vol. 51   page: 00-00   2014

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  216. A new technique for spectral interference correction on pin-by-pin BWR core analysis Reviewed

    T. Fujita,T. Endo,A. Yamamoto

    J. Nucl. Sci. Technol.   Vol. 51   page: 783-797   2014

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  217. Cross Section Adjustment Method Based on Random Sampling Technique Reviewed

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

    J. Nucl. Sci. Technol.   Vol. 51   page: 590-599   2014

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  218. A macroscopic cross-section model for BWR pin-by-pin core analysis Reviewed

    T. Fujita,T. Endo,A. Yamamoto

    J. Nucl. Sci. Technol.   Vol. 51   page: 282-304   2014

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  219. Application of Augmented Reality to Nuclear Reactor Core Simulation for Fundamental Nuclear Engineering Education Reviewed

    K. Tsujita,T. Endo,A. Yamamoto

    Nucl. Technol.   Vol. 185   page: 71-84   2014

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  220. Estimation of Self-shielding Effect on Uncertainty of Neutronics Characteristics using Random Sampling Method and Continuous-energy Slowing-down Calculation Reviewed

    A. Yamamoto,S. Sato,T. Endo

    Trans. Am. Nucl. Soc.   Vol. 109   page: 1436-1438   2013.11

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  221. Uncertainty and Correlation Estimation of Reload Safety Parameters of PWR using Random Sampling Method Reviewed

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

    Trans. Am. Nucl. Soc.   Vol. 109   page: 1365-1368   2013.11

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  222. Behavior of Higher Order Fission Source Distribution in Monte-Carlo Calculations Reviewed

    A. Yamamoto,K. Sakata,T. Endo

    Trans. Am. Nucl. Soc.   Vol. 109   page: 1361-1364   2013.11

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  223. Subcriticality Measurement Technique using Inherent Neutron Source in Uranium Fuel Reviewed

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

    Trans. Am. Nucl. Soc.   Vol. 109   page: 826-829   2013.11

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  224. Evaluation of Higher Order Mode Components of Fission Source Distribution in Monte Carlo Calculation Reviewed

    K. Sakata,T. Endo,A. Yamamoto

    Proc. Int. Conf. SNA&MC2013, Paris, France, Oct. 27-31, 2013     2013.10

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  225. Higher Order Treatment on Temporal Derivative of Angular Flux for Time-Dependent MOC Reviewed

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

    Proc. Int. Conf. Math. and Comp. Methods Applied to Nucl. Sci. Eng. (M&C2013)     2013.5

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  226. Explicit Estimation of Higher Order Modes in Fission Source Distribution of Monte-Carlo Calculation Reviewed

    A. Yamamoto,K. Sakata,T. Endo

    Proc. Int. Conf. Math. and Comp. Methods Applied to Nucl. Sci. Eng. (M&C2013)     2013.5

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  227. Reduction of Discretization Error for Ray Tracing of MOC Through a Correction on Collision Probabilities Reviewed

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

    Proc. Int. Conf. Math. and Comp. Methods Applied to Nucl. Sci. Eng. (M&C2013)     2013.5

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  228. Application of the Multigrid Amplitude Function Method for Time-Dependent Transport Equation Using MOC Reviewed

    K. Tsujita,T. Endo,A. Yamamoto

    Proc. Int. Conf. Math. and Comp. Methods Applied to Nucl. Sci. Eng. (M&C2013)     2013.5

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  229. Random Sampling-Based Cross-Section Adjustment Technique for LWR Core Analysis Reviewed

    S. Kato,T. Endo,A. Yamamoto

    Proc. ICAPP2013, Jeju, Korea Apr. 14-18, 2013     2013.4

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  230. Correction Technique for Course Group Cross Sections Considering Spectral Interference Effect on Pin-by-Pin BWR Core Analysis Reviewed

    T. Fujita,T. Endo,A. Yamamoto

    Proc. ICAPP2013, Jeju, Korea Apr. 14-18, 2013     2013.4

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  231. Convergence analysis of MOC inner iterations with large negative self-scattering cross-section Reviewed

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

    J. Nucl. Sci. Technol.   Vol. 50   page: 493-502   2013

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  232. Uncertainty estimation of core safety parameters using cross-correlations of covariance matrix Reviewed

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

    J. Nucl. Sci. Technol.   Vol. 50   page: 966-978   2013

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  233. Preservation of transmission probabilities in the method of characteristics Reviewed

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

    J. Nucl. Sci. Technol.   Vol. 50   page: 837-843   2013

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  234. Study on Discontinuity Factor for Angular Flux in Transport Equation Reviewed

    T. Sakamoto,T. Endo,A. Yamamoto

    Trans. Am. Nucl. Soc.   Vol. 108   page: 887-890   2013

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  235. Estimation of Self-shielding Effect on Uncertainty of Neutronics Characteristics using Random Sampling Method and Continuous-energy Slowing-down Calculation Reviewed

    A. Yamamoto,S. Sato,T. Endo

    Trans. Am. Nucl. Soc.   Vol. 109   page: 1436-1438   2013

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  236. Uncertainty and Correlation Estimation of Reload Safety Parameters of PWR using Random Sampling Method Reviewed

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

    Trans. Am. Nucl. Soc.   Vol. 109   page: 1365-1368   2013

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  237. Behavior of Higher Order Fission Source Distribution in Monte-Carlo Calculations Reviewed

    A. Yamamoto,K. Sakata,T. Endo

    Trans. Am. Nucl. Soc.   Vol. 109   page: 1361-1364   2013

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  238. Subcriticality Measurement Technique using Inherent Neutron Source in Uranium Fuel Reviewed

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

    Trans. Am. Nucl. Soc.   Vol. 109   page: 826-829   2013

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  239. Few-Group Macroscopic Cross Section Adjustment for LWRs Using Random Sampling Technique Reviewed

    A. Yamamoto,S. Kato,T. Endo

    Trans. Am. Nucl. Soc.   Vol. 108   page: 894-897   2013

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  240. Utilization of Discontinuity Factor in Integro-differential Type of Boltzmann Transport Equation Reviewed

    A. Yamamoto

    Nucl. Sci. Eng   Vol. 172   page: 259-267   2012.11

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  241. Development of Erbia Credit Super High Burnup Fuel: Evaluation of Minimum Erbia Content for Criticality Safety Analyses Reviewed

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

    Nucl. Technol.   Vol. 180   page: 18-27   2012.10

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

    Y. Tur,T. Endo,A. Yamamoto

    J. Nucl. Sci. Technol.   Vol. 49   page: 1028-1041   2012.10

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  243. An Optimization Approach to Establish an Appropriate Energy Group Structure for BWR Pin-by-Pin Core Analysis Reviewed

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

    J. Nucl. Sci. Technol.   Vol. 49   page: 689-707   2012.7

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  244. Advanced Resonance Self-Shielding Method for Gray Resonance Treatment in Lattice Physics Code GALAXY Reviewed

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

    J. Nucl. Sci. Technol.   Vol. 49   page: 725-747   2012.7

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  245. Efficient Fission Neutron Spectrum Matrix Representation by Singular Value Decomposition Technique Reviewed

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

    J. Nucl. Sci. Technol.   Vol. 49   page: 748-753   2012.7

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  246. Analysis of Erbia-Loaded Critical Experiments in KUCA Using AEGIS Cross Section Library Reviewed

    A. Yamamoto,T. Endo,X. Wu

    Trans. Am. Nucl. Soc.   Vol. 106   page: 715-718   2012.6

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  247. Kinetic Calculation Method in Space-Time Frame Using Characteristic Line Reviewed

    K. Tsujita,T. Endo,A. Yamamoto

    Trans. Am. Nucl. Soc.   Vol. 106   page: 743-746   2012.6

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  248. A Unified Approach for Numerical Calculation of Space-dependent Kinetics Equation Reviewed

    Y. Ban,T. Endo,A. Yamamoto

    J. Nucl. Sci. Technol.   Vol. 49   page: 496-515   2012.5

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  249. Uncertainty Estimation of Core Safety Parameters using Cross-Correlations of Covariance Matrix Reviewed

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

    Proc. Physor2012 - Advanced in Reactor Physics - Linking research, Industry and Education     2012.4

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  250. Estimation of average burnup of damaged fuels loaded in Fukushima Dai-ichi reactors by using the 134Cs/137Cs ratio method Reviewed

    T. Endo,S. Sato,A. Yamamoto

    Proc. Physor2012 - Advanced in Reactor Physics - Linking research, Industry and Education     2012.4

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  251. Multi-Physics Nuclear Reactor Simulator for Advanced Nuclear Engineering Education Reviewed

    A. Yamamoto

    Proc. Physor2012 - Advanced in Reactor Physics - Linking research, Industry and Education     2012.4

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  252. Development of Erbia Credit Super High Burnup Fuel: Experiments and Numerical Analyses Reviewed

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

    Nucl. Technol.   Vol. 177   page: 63-72   2012.1

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  253. Correction of Spectral Interference Effect on Pin-by-Pin BWR Core Analysis Reviewed

    T. Fujita,T. Endo,A. Yamamoto

    Trans. Am. Nucl. Soc.   Vol. 107   page: 1141-1143   2012

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  254. Efficient Calculation Scheme with Preservation of Transmission Probabilities in the Method of Characteristics Reviewed

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

    Trans. Am. Nucl. Soc.   Vol. 107   page: 1105-1107   2012

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  255. Higher order Treatment on Temporal Derivative of Angular Flux for Time-dependent MOC Reviewed

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

    Trans. Am. Nucl. Soc.   Vol. 107   page: 1101-1104   2012

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  256. Subcriticality Measurement by Neutron Source Multiplication Method with Detected-Neutron Multiplication Factor Reviewed

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

    Trans. Am. Nucl. Soc.,   Vol. 107   page: 648-651   2012

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  257. Detected-Neutron Multiplication Factor Measured by Neutron Source Multiplication Method Reviewed

    T. Endo,A. Yamamoto,Y. Yamane

    Ann. Nucl. Energy   Vol. 38   page: 2417-2427   2011.11

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  258. Evaluation of Correlation Among Uncertainties of Core Neutronic Parameters in LWR Reviewed

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

    Trans. Am. Nucl. Soc.   Vol. 105   page: 486-488   2011.11

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  259. Assembly Discontinuity Factor for Angular Flux in Transport Calculation Reviewed

    A. Yamamoto,T. Endo

    Trans. Am. Nucl. Soc.   Vol. 105   page: 862-864   2011.11

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  260. Development of a Lattice Physics Code for Sensitivity Analysis Based on Generalized Perturbation Theory Reviewed

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

    Trans. Am. Nucl. Soc.   Vol. 105   page: 851-854   2011.11

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  261. Application of the Discrete Ordinate CIP Scheme and the Fictitious Source Method for Reactor Shielding Reviewed

    K. Nakano, T. Endo, A. Yamamoto

    Proc. 19th International Conference On Nuclear Engineering (ICONE19)     page: 0   2011.10

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  262. A Systematic Approach to an Establish Energy Group Structure for BWR Pin-by-Pin Core Analysis Reviewed

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

    Proc. 19th International Conference On Nuclear Engineering (ICONE19)     page: 0   2011.10

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  263. Treatment of History Effect on Macroscopic Cross Section Model for BWR Pin-by-Pin Core Analysis Reviewed

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

    Proc. 19th International Conference On Nuclear Engineering (ICONE19)     page: 0   2011.10

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  264. A new calculation method for the generalized adjoint flux using the method of characteristics Reviewed

    S. Kato, T. Endo, A. Yamamoto

    Proc. 19th International Conference On Nuclear Engineering (ICONE19)     page: 0   2011.10

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  265. Application of Augmented Reality for Reactor Core Simulation Reviewed

    K. Tsujita, T. Endo, A. Yamamoto

    Proc. 19th International Conference On Nuclear Engineering (ICONE19)     page: 0   2011.10

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  266. Optimum In-Core Power Sharing with Multicycle Coupling Effect Reviewed

    A. Yamamoto,T. Iwata,Y. Yamane

    Prog. Nucl. Energy   Vol. 53   page: 593-599   2011.8

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  267. Application of the Robust Design Concept for Fuel Loading Pattern," Reviewed

    T. Endo,K. Ohori,A. Yamamoto

    J. Nucl. Sci. Technol.   Vol. 48   page: 1077-1086   2011.7

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  268. Improved Derivation of Multigroup Effective Cross Section for Heterogeneous System by Equivalence Theory Reviewed

    A. Yamamoto,T. Endo,G. Chiba

    Nucl. Sci. Eng   Vol. 168   page: 75-92   2011.6

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  269. Convergence Analysis of MOC with Large Negative Self-Scattering Cross Section Reviewed

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

    Trans. Am. Nucl. Soc.   Vol. 104   page: 809-811   2011.6

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  270. A Unified Numerical Algorithm for Space-Dependent Kinetic Equation Reviewed

    T. Endo,Y. Ban,A. Yamamoto

    Trans. Am. Nucl. Soc.   Vol. 104   page: 865-867   2011.6

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  271. A Derivation of Discontinuity Factor for Angular Flux in Integro-Differential Transport Equation Reviewed

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

    Trans. Am. Nucl. Soc.   Vol. 104   page: 815-817   2011.6

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  272. Resonance Calculation for Large and Complicated Geometry Using Tone's Method by Incorporating the Method of Characteristics Reviewed

    H. Yu,T. Endo,A. Yamamoto

    J. Nucl. Sci. Technol   Vol. 48   page: 330-336   2011.3

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  273. Overview of Core Simulation Methodologies for Light Water Reactor Analysis Reviewed

    A. Yamamoto,T. Endo

    Int. J. Nucl. Safety and Simulation     page: 12-21   2011.3

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  274. Explicit Time Integration Scheme using Krylov Subspace Method for Reactor Kinetics Equation Reviewed

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

    J. Nucl. Sci. Technol.   Vol. 48   page: 243-255   2011.2

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  275. Improvement of Tone's Method with Two-term Rational Approximation Reviewed

    A. Yamamoto,T. Endo,G. Chiba

    J. Nucl. Sci. Technol.   Vol. 48   page: 263-271   2011.2

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  276. Application of Quick Subchannel Analysis Method for Three-Dimensional Pin-by-Pin BWR Core Calculations Reviewed

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

    J. Nucl. Sci. Technol   Vol. 48   page: 1437-1452   2011

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  277. AEGIS: An Advanced Lattice Physics Code for Light Water Reactor Analyses Reviewed

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

    Nuclear Engineering and Technology   Vol. 42 ( 5 ) page: 500-519   2010.10

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  278. Incorporation of two-term rational approximation in Tone method for resonance calculation Reviewed

    A. Yamamoto,T. Endo,G. Chiba

    Trans. Am. Nucl. Soc.   Vol. 103   page: 711-713   2010.6

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  279. Investigation of theoretical approach to establish energy group structure for BWR pin-by-pin core analysis Reviewed

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

    Trans. Am. Nucl. Soc.   Vol. 103   page: 721-723   2010.6

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  280. Effect of uncertainty of planned cycle length in multi-cycle fuel optimization Reviewed

    K. Ohori,T. Endo,A. Yamamoto

    Trans. Am. Nucl. Soc.   Vol. 103   page: 707-710   2010.6

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  281. The Study on Erbia Credit Super-High-Burnup Fuel with Isotopically Modified Erbia Reviewed

    M. Yamasaki,H. Unesaki,A. Yamamoto

    Trans. Am. Nucl. Soc.   Vol. 103   page: 735-736   2010.6

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  282. Utilization of Discontinuity Factor in Integro-differential Type of Boltzmann Transport Equation Reviewed

    A. Yamamoto

    Proc. PHYSOR 2010 - Advances in Reactor Physics to Power the Nuclear Renaissance, Pittsburg, PA, May 2010   Vol. 1   2010.5

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  283. Resonance Calculation for Large and Complicated Geometry using Tone's Method by Incorporating Method of Characteristics Reviewed

    H. Yu,A. Yamamoto,Y. Yamane

    Proceedings of the 18th International Conference on Nuclear Engineering, ICONE18, Xian, China, May 2010   Vol. 1   2010.5

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  284. A New Robust Cross Section Representation Methodology for PWR Core Simulator Reviewed

    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   Vol. 1   2010.5

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  285. A Resonance Calculation Method based on the Multi-Terms Rational Approximation for General Geometry with Gray Resonance Absorbers Reviewed

    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   Vol. 1   2010.5

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  286. Numerical Solution of Reactor Kinetics Equation with Krylov Subspace Method for Matrix Exponential Reviewed

    Y. Ban,A. Yamamoto,Y. Yamane

    Proc. PHYSOR 2010 - Advances in Reactor Physics to Power the Nuclear Renaissance, Pittsburg, PA, May 2010   Vol. 1   2010.5

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  287. Investigation on Macroscopic Cross section Model for BWR Pin-by-pin Core Analysis Reviewed

    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   Vol. 1   2010.5

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  288. Validation of Neutron Current Formulations for the Response Matrix Method based on the SP3 Theory Reviewed

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

    Ann. Nucl. Energy   Vol. 37   page: 22-27   2010.1

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  289. 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   Vol. 46   page: 1099-1112   2009.12

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  290. An Improved Inverse Analysis Model for Fuel Loading Pattern Optimization Reviewed

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

      Vol. 46   page: 1162-1169   2009.12

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  291. *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   Vol. 100   page: 647-649   2009.11

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  292. Examination of Pin-by-pin Fission Rate Distribution in Large Geometry Evaluated by the Monte-Carlo Method Reviewed

    A. Yamamoto,R. Nakamura

    Ann. Nucl. Energy   Vol. 36   page: 1726-1733   2009.11

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  293. The Resonance Treatment in the AEGIS code

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

    Trans. Am. Nucl. Soc   Vol. 100   page: 654-655   2009.11

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  294. 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   Vol. 100   page: 650-651   2009.11

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  295. Projected Predictor-corrector Method for Lattice Physics Burnup Calculations Reviewed

    A. Yamamoto,M. Tatsumi,N. Sugimura

    Nucl. Sci. Eng   Vol. 163   page: 144-151   2009.10

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

    A. Yamamoto

    Trans. Am. Nucl. Soc.   Vol. 101   page: 402-404   2009.6

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  297. Subcriticality Estimation of Large FBR by Detectable Multiplication Factor kdet Reviewed

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

    Trans. Am. Nucl. Soc.   Vol. 101   page: 743-745   2009.6

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  298. 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.   Vol. 101   page: 737-738   2009.6

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  299. 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.   Vol. 101   page: 717-719   2009.6

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  300. 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   Vol. 101   page: 730-732   2009.6

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  301. 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)   Vol. CD-ROM   2009.5

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  302. Progress of Criticality Experiments and Nuclear Design Studies on Erbia-Bearing Super High Burnup Fuel Reviewed

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

    Proc. International Congress on Advances in Nuclear Power Plants (ICAPP09),   Vol. CD-ROM   2009.5

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  303. 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)   Vol. CD-ROM   2009.5

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  304. 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)   Vol. CD-ROM   2009.5

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  305. 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)   Vol. CD-ROM   2009.4

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  306. 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)   Vol. CD-ROM   2009.4

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

    A. Yamamoto

    Ann. Nucl. Energy   Vol. 36   page: 193-198   2009.3

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

    K. Tada,A. Yamamoto,Y. Yamane

    J. Nucl. Sci. Technol   Vol. 46   page: 163-174   2009.2

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  309. *Evaluation of Background Cross-section for Heterogeneous and Complicated Geometry by the Enhanced Neutron Current Method Reviewed

    A. Yamamoto

    J. Nucl. Sci. Technol   Vol. 45   page: 1287-1292   2008.12

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  310. Applicability of the Diffusion and the Simplified P3 Theories for Pin-by-pin Geometry of BWR Reviewed

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

    J. Nucl. Sci. Technol   Vol. 45   page: 997-1008   2008.10

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

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

    Proc. 16th Pacific Basin Nuclear Conference (16PBNC)   Vol. CD-ROM   2008.10

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  312. Some properties of zero power neutron noise in a time-varying medium with delayed neutrons Reviewed

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

    Ann. Nucl. Energy   Vol. 35   page: 1621-1627   2008.9

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  313. Development of a Prototype Pin-by-pin Fine Mesh Calculation Code for BWR Core Analysis Reviewed

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

    Proc. International Conference on the Physics of Reactors (Physor2008)     page: CD-ROM   2008.9

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  314. Verification of the AEGIS/SCOPE2 In-core Fuel Management System Reviewed

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

    Proc. International Conference on the Physics of Reactors (Physor2008)     page: CD-ROM   2008.9

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  315. Projected Predictor-corrector Method for Burnup Calculations of Gd-Bearing Fuel Assemblies Reviewed

    A. YamamotoM. Tatsumi,N. Sugimura

    Proc. International Conference on the Physics of Reactors (Physor2008)     page: CD-ROM   2008.9

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  316. Measuring the Ratio of 242Cm to 244Cm in Operating BWR Cores Using Start-Up Range Neutron Monitors Reviewed

    M. Watanabe,A. Yamamoto,Y. Yamane

    J. Nucl. Sci. Technol   Vol. 45   page: 498-509   2008.6

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  317. Fast computation of the Neutron Transport Calculation with a Graphic Processor Unit (GPU) Reviewed

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

    Trans. Am. Nucl. Soc   Vol. 99   page: 695-697   2008.6

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  318. Optimization of Batch Power Sharing to Improve Discharge Burnup for Multicycle Reviewed

    T. Iwata,A. Yamamoto,Y.Yamane

    Trans. Am. Nucl. Soc   Vol. 99   page: 703-705   2008.6

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  319. Development of a Resonance Calculation Method Based on Discrete Treatment of Energy Ranges Reviewed

    H. Koike,A. Yamamoto,Y.Yamane

    Trans. Am. Nucl. Soc   Vol. 99   page: 674-676   2008.6

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  320. Evaluation of the Background Cross Section for Heterogeneous and Complicated Geometry by the Enhanced Neutron Current Method Reviewed

    A. Yamamoto,

    Trans. Am. Nucl. Soc   Vol. 99   page: 671-673   2008.6

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  321. Application of a Game Console for Fast Reactor Physics Computation Reviewed

    Y. Kodama,A. Yamamoto,Y. Yamane

    Trans. Am. Nucl. Soc   Vol. 99   page: 698-699   2008.6

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  322. Reduction of Spatial Discretization Error for the Method of Characteristics using the Mobile-Chord Ray Tracing Method Reviewed

    A. Yamamoto

    Ann. Nucl. Energy   Vol. 35   page: 783-789   2008.5

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  323. Development of Deterministic Code Based on the Discrete Ordinates Method for the Third-Order Neutron Correlation Technique Reviewed

    T. Endo,A. Yamamoto,Y. Yamane

    Ann. Nucl. Energy   Vol. 35   page: 927-936   2008.5

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  324. Simplified Treatments of Anisotropic Scattering in LWR Core Calculations Reviewed

    A. YamamotoY. Kitamura,Y. Yamane

    J. Nucl. Sci. Technol   Vol. 45   page: 217-229   2008.3

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  325. Approximate Treatments of Anisotropic Scattering in LWR Analysis Reviewed

    A. Yamamoto

    Trans. Am. Nucl. Soc.   Vol. 96   page: 505-507   2007.11

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  326. Verification of Real-time Subcriticality Measurement based on Rossi-alpha Method Using Detection-time Acquisition System Reviewed

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

    Trans. Am. Nucl. Soc.   Vol. 96   page: 625-626   2007.11

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  327. Treatment of Staggered Mesh in BWR Pin-by-pin Fine Mesh Core Analysis Reviewed

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

    Trans. Am. Nucl. Soc.   Vol. 96   page: 508-510   2007.11

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  328. A Barrier on the Public Communication of Nuclear Technology - How to Interpret Reactor Kinetics Reviewed

    A. Yamamoto

    Proc. International Symposium on EcoTopia Science 2007     page: CD-ROM   2007.11

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  329. Development of New Solid Angle Quadrature Sets to Satisfy Even- and Odd-Moment Conditions Reviewed

    T. Endo, A. Yamamoto

    J. Nucl. Sci. Technol.   Vol. 44   page: 1249-1258   2007.10

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  330. Resonance Treatment Based on Ultra-fine-group Spectrum Calculation in the AEGIS Code Reviewed

    N. Sugimura, A. Yamamoto

    J. Nucl. Sci. Technol.   Vol. 44   page: 958-966   2007.7

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  331. Accuracy of a Rapid Cell-Heterogeneous Calculation Method for LWR Core Analysis Reviewed

    A. Yamamoto

    Trans. Am. Nucl. Soc.   Vol. 97   page: 582-584   2007.6

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  332. Development of Er-SHB fuel: Critical Experiments and Analyses of Homogeneously Erbia-Loaded Cores in KUCA Reviewed

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

    Trans. Am. Nucl. Soc.   Vol. 97   page: 715-717   2007.6

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  333. Applicability of the SP3 Nodal Method for BWR Pin-by-pin Core Analysis with Staggered Mesh Reviewed

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

    Trans. Am. Nucl. Soc.   Vol. 97   page: 569-572   2007.6

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  334. AEGIS/SCOPE2, a Next-Generation In-core Fuel Management System (2) Core Calculation Code, SCOPE2 Reviewed

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

    Trans. Am. Nucl. Soc   Vol. 97   page: 562-564   2007.6

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  335. Development of a New Lattice Physics Code GALAXY for Flexible Geometry Representation in NextGeneration Core Analyses System Reviewed

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

    Trans. Am. Nucl. Soc   Vol. 97   page: 573-574   2007.6

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  336. AEGIS/SCOPE2, a Next-Generation In-core Fuel Management System (1) Lattice Physics Code, AEGIS Reviewed

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

    Trans. Am. Nucl. Soc   Vol. 97   page: 559-561   2007.6

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  337. Calculation of higher moments of the neutron multiplication process in a time-varying medium Reviewed

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

    Ann. Nucl. Energy   Vol. 34   page: 385-395   2007.5

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  338. Erbia-bearing Super High Burnup Fuel: A Pathway for Breaking 5wt% Enrichment Barrier in LWR Fuel Reviewed

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

    Proc. International Conference on Nuclear Engineering, ICONE15   Vol. CD-ROM   2007.4

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  339. Performance of the Diffusion and Simplified PN Theories For BWR Pin-By-Pin Fine Mesh Core Analyses Reviewed

    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)   Vol. CD-ROM   2007.4

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  340. Applicability of The Diffusion and Simplified P3 Theories for BWR Pin-by-pin Core Analysis Reviewed

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

    Proc. International Conference on Nuclear Engineering, ICONE15   Vol. CD-ROM   2007.4

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  341. Fuel Loading Pattern Optimization Based on Case-Based Reasoning (CBR) Reviewed

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

    Proc. International Conference on Nuclear Engineering, ICONE15   Vol. CD-ROM   2007.4

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  342. Evaluation of Core Characteristics of FBR and PWR Loaded with Americium-containing Cermet Fuel Reviewed

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

    Proc. International Conference on Nuclear Engineering, ICONE15   Vol. CD-ROM   2007.4

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  343. Application of the Mobile-Chord Method for the Method of Characteristics Reviewed

    A. Yamamoto

    Proc. International Conference on Nuclear Engineering, ICONE15   Vol. CD-ROM   2007.4

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  344. Neutron Transport Models of AEGIS: An Advanced Next-Generation Neutronics Design System Invited Reviewed

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

    Nucl. Sci. Eng.   Vol. 155   page: 276-289   2007.2

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  345. Numerical Solution of Stiff Burnup Equation with Short Half Lived Nuclides by the Krylov Subspace Method Reviewed

    A. Yamamoto, M. Tatsumi, N. Sugimura

    J. Nucl. Sci. Technol.   Vol. 44   page: 147-154   2007.2

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  346. *Derivation of Optimum Polar Angle Quadrature Set for the Method of Characteristics Based on Approximation Error of the Bickley Function Reviewed

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

    J. Nucl. Sci. Technol.   Vol. 44   page: 129-136   2007.2

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  347. Improvement of Spatial Discretization Error on the Semi-analytic Nodal Method using the Scattered Source Subtraction Method Reviewed

    A. Yamamoto, M. Tatsumi

    J. Nucl. Sci. Technol.   Vol. 43   page: 1481-1489   2006.12

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  348. Reduction of the Spatial Discretization Error in the Method of Characteristics using the Diamond-difference Scheme Reviewed

    A. Yamamoto

    J. Nucl. Sci. Technol.   Vol. 43   page: 1327-1335   2006.11

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  349. Effect of error propagation of nuclide number densities on Monte Carlo burn-up calculations Reviewed

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

    Ann. Nucl. Energy   Vol. 33   page: 1424-1436   2006.11

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  350. Evaluation of Dancoff Factors in Complicated Geometry Using the Method of Characteristics Reviewed

    N. Sugimura, A. Yamamoto

    J. Nucl. Sci. Technol.   Vol. 43   page: 1182-1187   2006.10

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  351. Generation of Cross Section Library for Lattice Physics Code, AEGIS Reviewed

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

    Proc. Physor-2006   Vol. CD-ROM   2006.9

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  352. Development of the New Pin-by-Pin Core Calculation Method with Embedded Heterogeneous Assembly Calculation Reviewed

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

    Proc. Physor-2006   Vol. CD-ROM   2006.9

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  353. Verification of AEGIS/SCOPE2, a Next-Generation In-Core Fuel Management System Reviewed

    M. Tatsumi, N. Sugimura, A. Yamamoto

    Proc. Physor-2006   Vol. CD-ROM   2006.9

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  354. Calculation Models of AEGIS/SCOPE2, a Core Calculation System of Next Generation Reviewed

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

    Proc. Physor-2006   Vol. CD-ROM   2006.9

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  355. Application of Pin-by-pin Fine Mesh Calculation Method to BWR Core Analysis Reviewed

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

    Proc. Physor-2006   Vol. CD-ROM   2006.9

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  356. Feynman-alpha experiment with stationary multiple emission sources Reviewed

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

    Progress in Nuclear Energy   Vol. 48   page: 569-577   2006.8

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  357. Application of Variance-to-mean Technique to Subcriticality Monitoring for Accelerator-driven Sub-critical Reactor Reviewed

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

    Int. J. of Nucl. Energy Sci. and Technol.   Vol. 2   page: 266-284   2006.7

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  358. Derivation of Theoretical Formula for the Third Order Neutron Correlation Technique by Using Importance Function Reviewed

    T. Endo, Y. Yamane, A. Yamamoto

    Ann. Nucl. Energy   Vol. 33   page: 857-868   2006.7

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  359. Application of the Krylov Subspace Method to Burnup Calculation for Lattice Physics Code Reviewed

    A. Yamamoto, M. Tatsumi, N. Sugimura

    Trans. Am. Nucl. Soc.   Vol. 95   page: 713-714   2006.6

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  360. Development of Erbia-bearing Super High Burnup Fuel Reviewed

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

    Proc. International Congress on Advances in Nuclear Power Plants, ICAPP06     page: 1874-1882   2006.6

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  361. Reduction in Workload of BWR In-Core Fuel Shuffling by New Optimization Methods Reviewed

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

    Nucl. Technol.   Vol. 154   page: 318-327   2006.6

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  362. Improvement on Multi-group Scattering Matrix in Thermal Energy Range Generated by NJOY Reviewed

    A. Yamamoto, N. Sugimura

    Ann. Nucl. Energy   Vol. 33   page: 555-559   2006.4

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  363. Space and Energy Dependent Theoretical Formula for the Third Order Neutron Correlation Technique Reviewed

    T. Endo, Y. Yamane, A. Yamamoto

    Ann. Nucl. Energy   Vol. 33   page: 521-537   2006.4

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  364. Three-dimensional Pin Power Reconstruction for the Axially Heterogeneous Region in BWR Reviewed

    M. Tohjoh,M. Watanabe,A. Yamamoto

    Ann. Nucl. Energy   Vol. 33   page: 242-251   2006.2

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  365. Study of the Spatial Discretization and Temperature Distribution Approximation Effects on BWR Assembly Calculations Reviewed

    M. Tohjoh, M. Watanabe, A. Yamamoto

    Ann. Nucl. Energy   Vol. 33   page: 170-179   2006.1

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  366. Calculation of the Stochastic Pulsed Rossi-alpha Formula and its Experimental Verification Reviewed

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

    Prog. Nucl. Energy   Vol. 48   page: 37-50   2006.1

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  367. Improvement of the Flat Source Approximation in the Method of Characteristics Reviewed

    A. Yamamoto

    Trans. Am. Nucl. Soc.   Vol. 95   page: 577-578   2006

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  368. Generalized Coarse Mesh Rebalance Method for Acceleration of Neutron Transport Calculations Reviewed

    A. Yamamoto

    Nucl. Sci. Eng.   Vol. 151   page: 274-282   2005.11

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  369. Verification Calculations of AEGIS, an Advanced Neutronics Solver of Next-Generation Reviewed

    N. Sugimura, T. Ushio, A. Yamamoto

    Trans. Am. Nucl. Soc.   Vol. 92   page: 633-634   2005.11

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  370. Calculation Models of AEGIS, an Advanced Neutronics Solver of Next-Generation Reviewed

    A. Yamamoto, N. Sugimura, T. Ushio

    Trans. Am. Nucl. Soc.   Vol. 92   page: 631-632   2005.11

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  371. A New Optimization Algorithm for In-core Fuel Shuffling Sequence of BWR Reviewed

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

    Trans. Am. Nucl. Soc.   Vol. 92   page: 605-606   2005.11

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  372. Effects of the spatial discritization and temperature distribution approximation on BWR assembly calculations Reviewed

    M. Tohjoh, M. Watanabe, A. Yamamoto

    Proc. International topical Meeting on Mathematics and Computation, Supercomputing, Reactor Physics and Nuclear and Biological Applications (M&C2005)   Vol. CD-ROM   2005.9

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  373. Medical Isotope Production Using Pressurized Water Reactor Reviewed

    T. Endo, A. Yamamoto

    Proc. International Symposium on Isotope Science and Engineering from Basic to Applications   Vol. CD-ROM   2005.9

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  374. Efficient Calculation of Time-Dependent Neutron Transport Equation Using The Constrained Interpolated Profile (CIP) Method Reviewed

    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)   Vol. CD-ROM   2005.9

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  375. Non-equidistant Ray Tracing for the Method of Characteristics Reviewed

    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)   Vol. CD-ROM   2005.9

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  376. Development of Advanced Neutronics Design System of Next Generation, AEGIS Reviewed

    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)   Vol. CD-ROM   2005.9

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  377. Cost Evaluation of Power Uprate due to Reduction of In-core Power Peaking Factor Reviewed

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

    Trans. Am. Nucl. Soc.   Vol. 93   page: 378-379   2005.6

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  378. Effect of Anisotropic Scattering in PWR/APWR Radial-Reflector Calculations Reviewed

    A. Yamamoto, N. Sugimura, T. Ushio

    Trans. Am. Nucl. Soc.   Vol. 93   page: 617-618   2005.6

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  379. Yet Another Optimum Polar Angle Quadrature Set for the Method of Characteristics Reviewed

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

    Trans. Am. Nucl. Soc.   Vol. 93   page: 506-507   2005.6

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  380. Application of continuous-energy Monte Carlo code as a cross-section generator of BWR core calculations Reviewed

    M. Tohjoh, M. Watanabe, A. Yamamoto

    Ann. Nucl. Energy   Vol. 32   page: 857-875   2005.5

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  381. Calculation of the Pulsed Feynman- and Rossi-alpha Formulae with Delayed Neutrons Reviewed

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

    Ann. Nucl. Energy   Vol. 32   page: 671-692   2005.5

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  382. Convergence Property of Response Matrix Method for Various Finite-Difference Formulations used in the Non-Linear Acceleration Method Reviewed

    A. Yamamoto

    Nucl. Sci. Eng.   Vol. 149   page: 259-269   2005.3

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  383. A Conceptual Design Study of Proliferation-Resistant PWR Fuel using Reprocessed Uranium Reviewed

    A. Yamamoto, Y. Kagagiri, Y. Yamane

    Nucl. Eng. Des.   Vol. 235   page: 649-660   2005.3

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  384. Impact of Pin-by-pin Thermal-hydraulic Feedback Modeling on Steady-state Core Characteristics Reviewed

    A. Yamamoto, T. Ikeno

    Nucl. Technol.   Vol. 149   page: 175-188   2005.2

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  385. Improvement of the SPH Method for Pin-by-pin Core Calculations Reviewed

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

    J. Nucl. Sci. Technol.   Vol. 41   page: 1155-1165   2004.12

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  386. Simultaneous Loading Patterns Optimization for Two Successive Cycles of Pressurized Water Reactors Reviewed

    A. Yamamoto, E. Sugimura, Y. Kitamura, Y. Yamane

    J. Nucl. Sci. Technol.   Vol. 41   page: 1065-1074   2004.11

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  387. Approximate Treatment of Thermal Expansion Effect in Lattice Transport Calculations Reviewed

    A. Yamamoto, Y. Kitamura, Y. Yamane

    J. Nucl. Sci. Technol.   Vol. 41   page: 1003-1007   2004.10

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  388. Convergence Improvement of Coarse Mesh Rebalance Method for Neutron Transport Calculations Reviewed

    A. Yamamoto, Y. Kitamura, T. Ushio, N. Sugimura

    J. Nucl. Sci. Technol.   Vol. 41   page: 781-789   2004.8

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  389. Acceleration of Response Matrix Method Using Cross Section Scaling Reviewed

    A. Yamamoto

    Nucl. Sci. Eng.   Vol. 147   page: 176-184   2004.6

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  390. Computational Efficiencies of Approximated Exponential Functions for Transport Calculations of the Characteristics Method Reviewed

    A. Yamamoto, Y. Kitamura, Y. Yamane

    Ann. Nucl. Energy   Vol. 31   page: 1027-1037   2004.6

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  391. Comparison of 2-Group/9-Group Nodal-Transport Calculations in 3-D Pin-by-Pin Geometry Reviewed

    M. Tatsumi, A. Yamamoto

    Trans. Am. Nucl. Soc.   Vol. 91   page: 264-247   2004.6

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  392. Comparison of Subcriticality Monitoring Methods for Accelerator-Driven System Reviewed

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

    Trans. Am. Nucl. Soc.   Vol. 91   page: 751-752   2004.6

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  393. Simultaneous In-core Optimization of PWR Tandem Cycles Reviewed

    A. Yamamoto, E. Sugimura, Y. Kitamura, Y. Yamane

    Trans. Am. Nucl. Soc.   Vol. 91   page: 766-767   2004.6

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  394. Cell Homogenization Methods for Pin-by-pin Core Calculations Tested in Slab Geometry Reviewed

    A. Yamamoto, Y. Kitamura, Y. Yamane

    Ann. Nucl. Energy   Vol. 31   page: 825-847   2004.5

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  395. Improvement of the SPH Method for Multi-assembly Calculations Reviewed

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

    Proc. The Physics of Fuel Cycles and Advanced Nuclear Systems: Global Developments (Physor2004)   Vol. CD-ROM   2004.4

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  396. Derivation of the Space and Energy Dependent Formula for the Third Order Neutron Correlation Technique Reviewed

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

    Proc. The Physics of Fuel Cycles and Advanced Nuclear Systems: Global Developments (Physor2004)   Vol. CD-ROM   2004.4

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  397. The feasibility study of the minimum-shuffling reloading strategy for PWR Reviewed

    M. Tabuchi, Y. Hanayama, M. Yamasaki, A. Yamamoto

    Proc. The Physics of Fuel Cycles and Advanced Nuclear Systems: Global Developments (Physor2004)   Vol. CD-ROM   2004.4

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  398. Sensitivity Analysis for Multiplication Factor Change of LWR Cell Caused by the Differences between JENDL3.2 and JENDL3.3 Reviewed

    K. Kitada, T. Takeda, M. Yamasaki, M. tatsumi, A. Yamamoto

    J. Nucl. Sci. Technol.   Vol. 41   page: 163-170   2004.2

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  399. A Simple and Efficient Control Rod Cusping Model for Three-Dimensional Pin-By-Pin Core Calculations Reviewed

    A. Yamamoto

    Nucl. Technol.   Vol. 145   page: 11-17   2004.1

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  400. Analytic Derivation of the Correction Factor for the Improved Coarse Mesh Method Reviewed

    A. Yamamoto

    Ann. Nucl. Energy   Vol. 31   page: 71-86   2004.1

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  401. Convergence Improvements of Response Matrix Method with Large Discontinuity Factors Reviewed

    A. Yamamoto

    Nucl. Sci. Eng.   Vol. 145   page: 291-298   2003.11

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  402. Application of Neural Network for Loading Pattern Screening of In-Core Optimization Calculations Reviewed

    A. Yamamoto

    Nucl. Technol.   Vol. 144   page: 63-75   2003.10

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  403. MERIT-Factor: A New Concept for Evaluation of the Economic Efficiency of Core Loading Patterns Reviewed

    M. Yamasaki, M. Yoshikuni, A. Yamamoto

    Proc. Advances in Nuclear Fuel Management III   Vol. CD-ROM   2003.10

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  404. Pin-by-pin Thermal-Hydraulic Feedback Modeling in Three-Dimensional Fine-Mesh Core Calculations Reviewed

    A. Yamamoto, T. Ikeno

    Proc. Advances in Nuclear Fuel Management III   Vol. CD-ROM   2003.10

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  405. Study on Neutronics Design of Accelerator Driven Subcritical Reactor as Future Neutron Source, Part1; Static Characteristics Reviewed

    A. Yamamoto, S. Shiroya

    Ann. Nucl. Energy   Vol. 30   page: 1409-1424   2003.9

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  406. Performance of a Fine-Grained Parallel Model for Multi-Group Nodal-Transport Calculations in Three-Dimensional Pin-By-Pin Reactor Geometry Reviewed

    M. Tatsumi, A. Yamamoto

    Proc. Int. Conf. on Supercomputing in Nuclear Applications (SNA2003)   Vol. CD-ROM   2003.9

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    Language:English  

    DOI: CD-ROM

  407. PWR Core Tracking Using a Next-Generation Core Calculation Code, SCOPE2 Reviewed

    M. Tatsumi, A. Yamamoto, H. Nagano, K. Sengoku

    Proc. Int.l Conf. on Global Environment and Advanced Nuclear Power Plants (GENES4/ANP2003)   Vol. CD-ROM   2003.9

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    Language:English  

    DOI: CD-ROM

  408. Study on Neutronics Design of Accelerator Driven Subcritical Reactor as Future Neutron Source, Part2; Kinetic Characteristics Reviewed

    A. Yamamoto, S. Shiroya

    Ann. Nucl. Energy   Vol. 30   page: 1425-1435   2003.9

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    Authorship:Lead author   Language:English   Publishing type:Research paper (scientific journal)  

  409. *Advanced PWR Core Calculation Based on Multi-group Nodal-transport Method in Three-dimensional Pin-by-Pin Geometry Reviewed

    M. Tatsumi, A. Yamamoto

    J. Nucl. Sci. Technol.   Vol. 40   page: 376-387   2003.6

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  410. Application of the Distributed Genetic Algorithm for In-Core Fuel Optimization Problems under Parallel Computational Environment Reviewed

    A. Yamamoto, H. Hashimoto

    J. Nucl. Sci. Technol.   Vol. 39   page: 1281-1288   2002.12

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  411. Effect of Radial Peaking Factor Limitation on Discharge Burnup Reviewed

    A. Yamamoto

    J. Nucl. Sci. Technol.   Vol. 39   page: 1260-1268   2002.12

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  412. Acceleration of Response Matrix Method by Cross Section Scaling Reviewed

    A. Yamamoto

    Proc. International Conference on the New Frontiers of Nuclear Technology: Reactor Physics, Safety and High-Performance Computing     page: 1A-03   2002.10

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  413. SCOPE2:Object-Oriented Parallel Code for Multi-Group Diffusion/Transport Calculations in Three-Dimensional Fine-Mesh Reactor Core Geometry Reviewed

    M. Tatsumi, A. Yamamoto

    Proc. International Conference on the New Frontiers of Nuclear Technology: Reactor Physics, Safety and High-Performance Computing     page: 12A-01   2002.10

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  414. Benchmark Problem Suite for Reactor Physics Study of LWR Next Generation Fuels Reviewed

    A. Yamamoto, T. Ikehara, T. Ito, E. Saji

    J. Nucl. Sci. Technol.   Vol. 39   page: 900-912   2002.8

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  415. Object-Oriented Three-Dimensional Fine-Mesh Transport Calculation on Parallel/Distributed Environments for Advanced Reactor Core Analyses Reviewed

    M. Tatsumi, A. Yamamoto

    Nucl. Sci. Eng.   Vol. 141   page: 190-217   2002.7

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    Language:English   Publishing type:Research paper (scientific journal)  

  416. PWR Core Calculations by the Multigroup Nodal-Transport Method in 3-D Pin-by-Pin Geometry Reviewed

    M. Tatsumi, A. Yamamoto

    Trans. Am. Nucl. Soc.   Vol. 87   page: 232-234   2002.6

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  417. Basic study on neutronics of future neutron source based on accelerator driven subcritical reactor concept in Kyoto University Research Reactor Institute (KURRI) Reviewed

    S. Shiroya, A. Yamamoto, K. Shin, T. Ikeda, S. Nakano, H. Unesaki

    Prog. Nucl. Energy   Vol. 40   page: 489-496   2002.3

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  418. Recent Activities of Loading Pattern Optimization Research in Japan Invited Reviewed

    A. Yamamoto

    Trans. Am. Nucl. Soc   Vol. 84   page: 57-59   2001.11

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  419. A Conceptual Neuronics Design Study for Next Generation Neutron Source in Kyoto University research Reactor Institute (KURRI) Reviewed

    A. Yamamoto, S. Shiroya

    Proc. ICENES 2000: The Tenth International Conference on Emerging Nuclear Energy Systems, Petten   Vol. CD-ROM   page: 66-75   2000.9

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  420. Effect of Core Calculation Models on Optimum Cycle Length Analyses of Pressurized Water Reactors Reviewed

    A. Yamamoto, T. Kimoto

    Ann. Nucl. Energy   Vol. 27   page: 1039-1050   2000.7

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  421. Analysis of the Ikata-3 Initial Core with the CHAPLET Heterogeneous Transport Calculation Code Based on the Method of Characteristics Invited Reviewed

    M. Tatsumi, T.Kimoto, A. Yamamoto

    Trans. Am. Nucl. Soc.   Vol. 83   page: 286-287   2000.6

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  422. Verification of Cell-Homogenized Whole-Core Transport Calculations in Actual PWR Core Geometry Invited Reviewed

    A. Yamamoto, M. Tatsumi, T. Kimoto, S, Kosaka, E. Saji

    Trans. Am. Nucl. Soc.   Vol. 83   page: 287-289   2000.6

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  423. Effect of Core Calculation Accuracy on Fuel Cycle Cost Reviewed

    A. Yamamoto

    Proc. Advances in Reactor Physics and Mathematics and Computation into the Next Millennium     page: Ⅶ.A-2   2000.5

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  424. A Study on Effects of Pin Cell Homogenization in an Actual Reactor Core Geometry Reviewed

    M. Tatsumi, A. Yamamoto,S.Kosaka,E.Saji

    Proc. Advances in Reactor Physics and Mathematics and Computation into the Next Millennium     page: IX.D-5   2000.5

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  425. A Study on Non-Proliferative, Retrofittable PWR fuel Reviewed

    Y. Katagiri, Y. Yamane, A. Yamamoto

    Proc. of ICONE 8: 8th International Conference on Nuclear Engineering   Vol. CD-ROM   page: 8366   2000.4

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  426. Application of Temperature Parallel Simulated Annealing to Loading Pattern Optimizations of Pressurized Water Reactors," Reviewed

    A. Yamamoto, H. Hashimoto

    Nucl Sci. Eng.   Vol. 136   page: 247-257   2000

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  427. Loading Pattern Optimizations in a Distributed Parallel Environment Invited Reviewed

    A. Yamamoto, H. Hashimoto

    Trans. Am. Nucl. Soc.   Vol. 80   page: 223-224   1999.11

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  428. Applications of Temperature Parallel Simulated Annealing to Loading Pattern Optimizations of Pressurized Water Reactors Invited Reviewed

    A. Yamamoto, H. Hashimoto

    Proc. Mathematics and Computation, Reactor Physics and Environmental Analysis in Nuclear Applications   Vol. 2   page: 1445-1458   1999.9

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  429. Advanced Reactor Core Analysis by the Object-oriented 3-D Fine Mesh Transport Calculation on Parallel/Distributed Environment Invited Reviewed

    M. Tatsumi, A. Yamamoto, H. Hashimoto

    Proc. Mathematics and Computation, Reactor Physics and Environmental Analysis in Nuclear Applications   Vol. 2   page: 1288-1297   1999.9

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  430. Effective Utilization of Weapon-Grade Plutonium to Upgrade Repeatedly-Reprocessed Mixed-Oxide Fuel for Use in Pressurized Water Reactors Reviewed

    Y. Hanayama, A. Yamamoto K. Kanda

    J. Nucl. Sci. Technol.   Vol. 36   page: 746-754   1999.9

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  431. A Comparison Between a One Point Reactor Model And a Two-Dimensional Core Model on Equilibrium Cycle Analysis of Pressurized Water Reactors Reviewed

    A. Yamamoto, T. Kimoto

    Proc. International Conference of the Physics of Nuclear Science and Technology   Vol. 1   page: 84-90   1998.10

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  432. Object-Oriented Approach for an Iterative Calculation Method and Its Parallelization with Domain Decomposition Method Reviewed

    M. Tatsumi, A. Yamamoto

    Lecture Notes in Computer Science   Vol. 1505   page: 1-12   1998.4

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  433. Development of Perturbation Code Based on Modified Explicit Higher Order Perturbation Method with Two-Energy-Group Reviewed

    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   Vol. 2   page: 1149-1158   1997.10

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  434. Comparison between Equilibrium Cycle and Successive Multicycle Optimization Methods for In-Core Fuel Management of Pressurized Water Reactors Reviewed

    A. Yamamoto

    Proc. Joint International Conference on Mathematical Methods and Supercomputing for Nuclear Applications   Vol. 1   page: 769-781   1997.10

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  435. SCOPE:A Scalable and Flexible Parallel Algorithm Based on Object-Oriented Approach for Core Calculations Reviewed

    M. Tatsumi, A. Yamamoto

    Proc. Joint International Conference on Mathematical Methods and Supercomputing for Nuclear Applications   Vol. 1   page: 191-202   1997.10

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  436. *Comparison between Equilibrium Cycle and Successive Multicycle Optimization Methods for In-Core Fuel Management of Pressurized Water Reactors Reviewed

    A. Yamamoto, K. Kanda

    J. Nucl. Sci. and Technol.   Vol. 34   page: 882-892   1997.9

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  437. INSIGHT: An Integrated Scoping Analysis Tool for In-Core Fuel Management of PWR Reviewed

    A. Yamamoto, H. Noda, T. Maruyama, N. Ito

    J. Nucl. Sci. and Technol.   Vol. 34   page: 847-855   1997.8

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  438. A Quantitative Comparison of Loading Pattern Optimization Methods for In-Core Fuel Management of PWR Reviewed

    A. Yamamoto

    J. Nucl. Sci. and Technol.   Vol. 34   page: 339-347   1997.4

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  439. Resonance Calculations Using the Multiband Method and Interface Current Reviewed

    M. Tatsumi, T. Ito, T. Takeda, M. Yamasaki, A. Yamamoto, M. Takayasu

    Nucl. Sci. Eng.   Vol. 125   page: 178-187   1997.3

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  440. INSIGHT: An Integrated Scoping Analysis Tool for In-Core Fuel Management of PWR Reviewed

    A. Yamamoto, H. Noda, N. Ito, T. Maruyama

    Proc. Advances in Nuclear Fuel Management II   Vol. 1   page: 8-1-8-11   1997.3

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  441. Development of Two-Energy Groups Higher Order Perturbation Method for Application to Core Analysis Reviewed

    C. H. Pyeon, Y. Yamane, T. Misawa, A. Yamamoto, K. Kagehira

    Proc. Advances in Nuclear Fuel Management II   Vol. 2   page: 15-11-15-18   1997.3

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  442. Core Burnup Calculations Using Neural Networks Reviewed

    H. Noda, A. Yamamoto, Y. Nagasawa, H. Murao, S. Kitamura,

    Proc. Advances in Nuclear Fuel Management II   Vol. 2   page: 20-39-20-50   1997.3

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  443. Transuranium Fuel Assembly for Transmutation in a Pressurized Water Reactor Invited Reviewed

    M. Mori, M. Kawamura, A. Yamamoto

    Nucl. Technol.   Vol. 117   page: 171-185   1997.2

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  444. Loading Pattern Optimizations Using Genetic Algorithms Reviewed

    A. Yamamoto

    Proc. International Conference on the Physics of Reactors (PHYSOR96)   Vol. 3   page: I-48-I-56   1996.9

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  445. Application of Nodal Method to Lambda Mode Higher Harmonics Code Reviewed

    T. Misawa, A. Yamamoto, Y. Yamane

    J. Nucl. Sci. Technol.   Vol. 33   page: 668-670   1996.8

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  446. Finite Difference Solution for Multigroup Transport Equation in R-Z Geometry by Spherical Harmonics Method Reviewed

    A. Yamamoto, K. Kobayashi

    J. Nucl. Sci. Technol.   Vol. 26   page: 563-574   1989.6

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▼display all

Books 3

  1. 原子炉の物理

    山本章夫、他( Role: Joint author)

    日本原子力学会  2019.12  ( ISBN:978-4-89047-172-0

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    Language:Japanese Book type:Textbook, survey, introduction

  2. Handbook of Nuclear Engineering

    Akio Yamamoto, et al.( Role: Joint author)

    Springer  2010 

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    Language:English

  3. プルトニウム燃料工学

    山本章夫、他( Role: Joint author)

    1998 

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    Language:Japanese

Presentations 90

  1. Verification of Neutronics Analysis Method Using CBZ and GENESIS for a Prismatic High-Temperature Gas-Cooled Reactor International conference

    山本章夫

    International Conference on Physics of Reactors 2024  2024.4.24  米国原子力学会

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    Event date: 2024.4

    Language:English   Presentation type:Oral presentation (general)  

    Venue:サンフランシスコ   Country:United States  

  2. Simplified Treatment of Coating Layers in TRISO Fuel in Statistical Geometry Method in Monte Carlo Calculation International conference

    山本章夫

    International Conference on Physics of Reactors 2024  2024.4.23  米国原子力学会

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    Event date: 2024.4

    Language:English   Presentation type:Oral presentation (general)  

    Venue:サンフランシスコ   Country:United States  

  3. CBZ-GENESISによるHTTR二次元炉心解析 (3) GENESISによる解析

    山本章夫, 千葉豪

    日本原子力学会 2024年春の年会  2024.3.27  日本原子力学会

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    Event date: 2024.3

    Language:Japanese   Presentation type:Oral presentation (general)  

    Venue:大阪   Country:Japan  

  4. Some Open Issues in Reactor Physics Simulations Invited International conference

    Akio Yamamoto

    M&C2023  2023.8.14  Canadian Nuclear Society

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    Event date: 2023.8

    Language:English   Presentation type:Oral presentation (keynote)  

    Venue:Niagara Falls, Canada   Country:Canada  

  5. Application of the RSE Method for the Resonance Treatment of HTGR Fuel with Double Heterogeneity International conference

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

    M&C2023  2023.8.14  Canadian Nuclear Society

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    Event date: 2023.8

    Language:English   Presentation type:Oral presentation (general)  

    Venue:Niagara Falls, Canada  

  6. Application of Equivalent Dancoff Factor Method for Resonance Calculation of Double Heterogeneous Fuel International conference

    Akio Yamamoto, Tomohiro Endo

    2023 ANS Annual Meeting  2023.6.14  American Nuclear Society

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    Event date: 2023.6

    Language:English   Presentation type:Oral presentation (general)  

    Venue:Indianapolis   Country:United States  

  7. RSE法を用いた二重非均質性を有する高温ガス炉燃料の共鳴計算

    ⼭本 章夫, 遠藤 知弘, 竹田 敏, 小池 啓基, 山路 和也, 浅野 耕司

    日本原子力学会 2023年春の年会  2023.3.14  日本原子力学会

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    Event date: 2023.3

    Language:Japanese   Presentation type:Oral presentation (general)  

    Venue:東京   Country:Japan  

  8. 核データ処理コードFRENDY 第2 版の開発:(2) 多群定数作成におけるresonance up-scattering の実装

    山本章夫,遠藤知弘,千葉豪,多田健一

    日本原子力学会 2022年秋の大会  2022.9.8  日本原子力学会

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    Event date: 2022.9

    Language:Japanese   Presentation type:Oral presentation (general)  

    Venue:日立市   Country:Japan  

  9. A New Approach for Resonance Treatment of Doubly Heterogeneous Fuel Using the RSE Method International conference

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

    International Conference on Physics of Reactors 2022 (PHYSOR 2022)   2022.5.18  American Nuclear Society

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    Event date: 2022.5

    Language:English   Presentation type:Oral presentation (general)  

    Venue:Pittsburgh   Country:United States  

  10. Safety assessments of advanced reactors Invited International conference

    Akio Yamamoto

    Asian Symposium on Risk Assessment and Management ASRAM2021  2021.10.26 

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    Event date: 2021.10

    Language:English   Presentation type:Symposium, workshop panel (nominated)  

    Venue:Online  

  11. FRENDY/MG: A Multi-group Cross Section Generation Module using ACE pointwise cross sections International conference

    M&C2021  2021.10.5  American Nuclear Society

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    Event date: 2021.10

    Language:English   Presentation type:Poster presentation  

  12. 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  2021.9.10  日本原子力学会

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    Event date: 2021.9

    Language:Japanese   Presentation type:Oral presentation (general)  

    Venue:オンライン  

  13. 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 Invited

    山本章夫

    日本原子力学会 2021年秋の大会  2021.9.10  日本原子力学会

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    Event date: 2021.9

    Language:Japanese   Presentation type:Oral presentation (invited, special)  

    Venue:オンライン  

  14. Verification of the Multi-Group Generation Capability of FRENDY Nuclear Data Processing Codefor Recent Nuclear Data through Comparison of One-group Reaction Rates International conference

    2021.6.15 

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    Event date: 2021.6

    Language:English   Presentation type:Oral presentation (general)  

  15. Contaminated Water Management: Current Situation and Issues in Fukushima-Daiichi Nuclear Power Station. Invited International conference

    The Management of Contaminated Water at Fukushima Daiichi  2021.4.21  National University of Singapore

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    Event date: 2021.4

    Language:English   Presentation type:Oral presentation (invited, special)  

  16. 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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    Event date: 2020.9

    Language:Japanese   Presentation type:Oral presentation (general)  

    Venue:オンライン   Country:Japan  

  17. 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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    Event date: 2020.9

    Language:Japanese   Presentation type:Oral presentation (general)  

    Venue:オンライン   Country:Japan  

  18. 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 Invited

    山本章夫

    日本機械学会2020年度年次大会 

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    Event date: 2020.9

    Language:Japanese   Presentation type:Oral presentation (general)  

    Venue:オンライン   Country:Japan  

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

    Yamamoto A.

    Transactions of the American Nuclear Society 

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    Event date: 2020.6

    Language:English   Presentation type:Oral presentation (general)  

    DOI: 10.13182/T122-32047

    Scopus

  20. Discontinuity Factor; A Discontinuity Condition for Angular Flux? International conference

    RPHA2019 

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    Event date: 2019.12

    Language:English   Presentation type:Oral presentation (general)  

    Country:Japan  

  21. Resonance Calculation Using Energy Spectral Expansion Based on Reduced Order Model: Application to Heterogeneous Geometry International conference

    ANS 2019 Winter Meeting 

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    Event date: 2019.11

    Language:English   Presentation type:Oral presentation (general)  

    Country:United States  

  22. Uncertainty Quantification/Reduction of BWR Core Characteristics Considering Cross Section and Thermal-Hydraulics Uncertainties International conference

    ANS 2019 Annual Meeting 

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    Event date: 2019.6

    Language:English   Presentation type:Oral presentation (general)  

    Country:United States  

  23. A Resonance Calculation Method using Energy Expansion Bases based on Reduced Order Model (1) Theory

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    Event date: 2019.3

    Language:Japanese   Presentation type:Oral presentation (general)  

    Country:Japan  

  24. Transport Consistent Diffusion Coefficient for CMFD Acceleration International conference

    A. Yamamoto, A. Giho, T. Endo

    ANS 2018 Winter Meeting 

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    Event date: 2018.11

    Language:English   Presentation type:Oral presentation (general)  

    Venue:Orlando, FL, USA   Country:United States  

  25. A Simple Treatment of Bowed Assembly Gap Through Correction of Cross Section International conference

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

    ANS 2018 Winter Meeting 

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    Event date: 2018.11

    Language:English   Presentation type:Oral presentation (general)  

    Venue:Orlando, FL, USA   Country:United States  

  26. Cache Efficient Flux Region Assignment for the Method of Characteristics International conference

    A. Yamamoto, A. Giho, T. Endo

    PHYSOR2018 

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    Event date: 2018.4

    Language:English   Presentation type:Oral presentation (general)  

    Venue:Cancun,Mexico   Country:Mexico  

  27. Inverse Estimation Methods of Unknown Radioactive Source for Fuel Debris Search International conference

    S. Sugaya, T. Endo, A. Yamamoto

    PHYSOR2018 

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    Event date: 2018.4

    Language:English   Presentation type:Oral presentation (general)  

    Venue:Cancun,Mexico   Country:Mexico  

  28. Estimation of Region-Wise Even-Parity Discontinuity Factor for MOC Through Iterative Procedure International conference

    A. Yamamoto, A. Giho, T. Endo

    PHYSOR2018 

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    Event date: 2018.4

    Language:English   Presentation type:Oral presentation (general)  

    Venue:Cancun,Mexico   Country:Mexico  

  29. 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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    Event date: 2018.3

    Language:Japanese   Presentation type:Oral presentation (general)  

    Venue:大阪大学吹田キャンパス   Country:Japan  

  30. Application of the GENESIS Code to the Kobayashi 3D Benchmark Problem International conference

    A. Yamamoto, A. Giho, T. Endo

    ANS 2017 Winter Meeting 

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    Event date: 2017.10 - 2017.11

    Language:English   Presentation type:Oral presentation (general)  

    Venue:Washington, D.C., USA   Country:United States  

  31. 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 

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    Event date: 2017.9

    Language:Japanese   Presentation type:Oral presentation (general)  

    Venue:北海道大学   Country:Japan  

  32. Development of GENESIS, a Three-dimensional Heterogeneous Transport Code based on the LEAF Method International conference

    A. Yamamoto, A. Giho, T. Endo

    M&C 2017 

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    Event date: 2017.4

    Language:English   Presentation type:Oral presentation (general)  

    Venue:Jeju, Korea   Country:Korea, Republic of  

  33. Application of Simplified Pn Approximation to Angular Distribution of Neutron Source in MOC Calculations International conference

    A. Yamamoto, A. Giho, T. Endo

    ANS 2016 Winter Meeting 

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    Event date: 2016.11

    Language:English   Presentation type:Oral presentation (general)  

    Venue:Las Vegas, Nevada, USA   Country:United States  

  34. Comparison of the Numerical Stability between CMFD and GCMR with Stabilization Techniques International conference

    A. Giho, A. Yamamoto, T. Endo

    ANS 2016 Winter Meeting 

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    Event date: 2016.11

    Language:English   Presentation type:Oral presentation (general)  

    Venue:Las Vegas, Nevada, USA   Country:United States  

  35. 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 

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    Event date: 2016.9

    Language:Japanese   Presentation type:Oral presentation (general)  

    Venue:久留米シティプラザ, 福岡県   Country:Japan  

  36. GENESIS - a Transport Solver in Three-Dimensional Heterogeneous Geometry based on the Leaf Method International conference

    A. Yamamoto, A. Giho, T. Endo

    PHYSOR2016 

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    Event date: 2016.5

    Language:English   Presentation type:Oral presentation (general)  

    Venue:Sun Valley, Idaho, USA   Country:United States  

  37. 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 

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    Event date: 2016.3

    Language:Japanese   Presentation type:Oral presentation (general)  

    Venue:東北大学川内キャンパス   Country:Japan  

  38. Discontinuity Factors for Simplified P3 Theory International conference

    A. Yamamoto, T. Sakamoto, T. Endo

    Reactor Physics Asia 2015 (RPHA15) 

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    Event date: 2015.9

    Language:English   Presentation type:Oral presentation (general)  

    Venue:Jeju, Korea   Country:Korea, Republic of  

  39. Reactor Physics Activities in Nagoya University International conference

    A. Yamamoto, T. Endo

    Reactor Physics Asia 2015 (RPHA15) 

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    Event date: 2015.9

    Language:English   Presentation type:Oral presentation (general)  

    Venue:Jeju, Korea   Country:Korea, Republic of  

  40. Angular Dependent Transmission Probability Method for Fast Reactor Core Transport Analysis International conference

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

    ANS 2015 Annual Meeting 

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    Event date: 2015.6

    Language:English   Presentation type:Oral presentation (general)  

    Venue:San Antonio, USA   Country:United States  

  41. 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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    Event date: 2014.3

    Language:Japanese   Presentation type:Oral presentation (general)  

    Venue:東京都市大学   Country:Japan  

  42. Estimation of Self-shielding Effect on Uncertainty of Neutronics Characteristicsusing Random Sampling Method and Continuous-energy Slowing-down Calculation International conference

    A. Yamamoto, S. Sato, T. Endo

    ANS 2013 Winter Meeting 

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    Event date: 2013.11

    Language:English   Presentation type:Oral presentation (general)  

    Venue:Washington D.C., USA   Country:United States  

  43. Behavior of Higher Order Fission Source Distribution in Monte-Carlo Calculations International conference

    A. Yamamoto, K. Sakata, T. Endo

    ANS 2013 Winter Meeting 

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    Event date: 2013.11

    Language:English   Presentation type:Oral presentation (general)  

    Venue:Washington D.C., USA   Country:United States  

  44. Few-Group Macroscopic Cross Section Adjustment for LWRs Using Random Sampling Technique International conference

    A. Yamamoto, S. Kato, T. Endo

    ANS 2013 Annual Meeting 

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    Event date: 2013.6

    Language:English   Presentation type:Oral presentation (general)  

    Venue:Atlanta, GA, USA   Country:United States  

  45. Explicit Estimation of Higher Order Modes in Fission Source Distribution of Monte-Carlo Calculation International conference

    A. Yamamoto, K. Sakata, T. Endo

    MC 2013 

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    Event date: 2013.5

    Language:English   Presentation type:Oral presentation (general)  

    Venue:Sun Valley, ID, USA   Country:United States  

  46. Application of the Multigrid Amplitude Function Method for Time-Dependent Transport Equation Using MOC International conference

    K. Tsujita, T. Endo, A. Yamamoto

    MC 2013 

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    Event date: 2013.5

    Language:English   Presentation type:Oral presentation (general)  

    Venue:Sun Valley, ID, USA   Country:United States  

  47. Higher Order Treatment on Temporal Derivative of Angular Flux for Time-Dependent MOC International conference

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

    MC 2013 

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    Event date: 2013.5

    Language:English   Presentation type:Oral presentation (general)  

    Venue:Sun Valley, ID, USA   Country:United States  

  48. 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 

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    Event date: 2013.3

    Language:Japanese   Presentation type:Oral presentation (general)  

    Venue:近畿大学東大阪キャンパス   Country:Japan  

  49. 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 

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    Event date: 2012.9

    Language:Japanese   Presentation type:Oral presentation (general)  

    Venue:広島大学東広島キャンパス   Country:Japan  

  50. Analysis of Erbia-Loaded Critical Experiments in KUCA Using AEGIS CrossSection Library International conference

    A. Yamamoto, T. Endo, X. Wu

    ANS 2012 Annual Meeting 

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    Event date: 2012.6

    Language:English   Presentation type:Oral presentation (general)  

    Venue:Chicago, Illinois,USA   Country:United States  

  51. Uncertainty Estimation of Core Safety Parameters Using Cross-Correlationsof Covariance Matrix International conference

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

    PHYSOR2012 

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    Event date: 2012.4

    Language:English   Presentation type:Oral presentation (general)  

    Venue:Knoxville, USA   Country:United States  

  52. Multi-Physics Nuclear Reactor Simulator for Advanced Nuclear Engineering Education International conference

    A Yamamoto

    PHYSOR2012 

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    Event date: 2012.4

    Language:English   Presentation type:Oral presentation (general)  

    Venue:Knoxville, USA   Country:United States  

  53. Assembly Discontinuity Factor for Angular Flux in Transport Calculation International conference

    A. Yamamoto, T. Endo

    ANS 2011 Winter Meeting 

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    Event date: 2011.10 - 2011.11

    Language:English   Presentation type:Oral presentation (general)  

    Venue:WashingtonD.C., USA   Country:United States  

  54. A Derivation of Discontinuity Factor for Angular Flux in Integro-Differential Transport Equation International conference

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

    ANS 2011 Annual Meeting 

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    Event date: 2011.6

    Language:English   Presentation type:Oral presentation (general)  

    Venue:Hollywood, Florida, USA   Country:United States  

  55. Incorporation of the Two-Term Rational Approximation for the Resonance Calculation with the Tone Method International conference

    A. Yamamoto, T. Endo, G. Chiba

    ANS 2010 Winter Meeting 

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    Event date: 2010.11

    Language:English   Presentation type:Oral presentation (general)  

    Country:United States  

  56. Improvement of Tone's method using the two-terms rational approximation

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    Event date: 2010.9

    Language:Japanese   Presentation type:Oral presentation (general)  

    Country:Japan  

  57. Utilization of Discontinuity Factor in Integro-differential Type of Boltzmann Transport Equation International conference

    PHYSOR 2010 

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    Event date: 2010.5

    Language:English  

  58. Application of discontinuity factor for integro-differential type transport equation

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    Event date: 2010.3

    Language:Japanese   Presentation type:Oral presentation (general)  

    Country:Japan  

  59. Application of Discontinuity Factor for Integro-Differential Transport Equation International conference

    ANS 2009 Winter meeting 

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    Event date: 2009.11

    Language:English   Presentation type:Oral presentation (general)  

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

    ANS 2009 Annual Meeting 

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    Event date: 2009.6

    Language:English  

  61. A New Framework of Resonance Calculation Method Based on the Sub-group Method (1); Theory International conference

    ANS 2009 Annual Meeting 

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    Event date: 2009.6

    Language:English  

  62. Optimum In-Core Power Sharing with Multi-cycle Coupling Effect International conference

    ANFM IV 

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    Event date: 2009.4

    Language:English  

  63. 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 

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    Event date: 2009.3

    Language:Japanese  

    Country:Japan  

  64. 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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    Event date: 2009.3

    Language:Japanese  

  65. Evaluation of the Background Cross Section for Heterogeneous and Complicated Geometry by the Enhanced Neutron Current Method International conference

    ANS 2008 Winter Meeting 

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    Event date: 2008.11

    Language:English  

  66. Projected Predictor-corrector Method for Burnup Calculations of Gd-Bearing Fuel Assemblies International conference

    PHYSOR 2008 

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    Event date: 2008.9

    Language:English  

  67. 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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    Event date: 2008.9

    Language:Japanese  

    Country:Japan  

  68. 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 

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    Event date: 2008.3

    Language:Japanese  

    Country:Japan  

  69. Development of Er-SHB fuel: Critical Experiments and Analyses of Homogeneously Erbia-Loaded Cores in KUCA International conference

    ANS 2007 Winter Meeting 

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    Event date: 2007.11

    Language:English  

  70. Accuracy of a Rapid Cell-Heterogeneous Calculation Method for LWR Core Analysis International conference

    ANS 2007 Winter Meeting 

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    Event date: 2007.11

    Language:English  

  71. Applicability of the SP3 Nodal Method for BWR Pin-by-pin Core Analysis with Staggered Mesh International conference

    ANS 2007 Winter Meeting 

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    Event date: 2007.11

    Language:English  

  72. Approximate Treatments of Anisotropic Scattering in LWR Analysis International conference

    ANS 2007 Annual Meeting 

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    Event date: 2007.6

    Language:English  

  73. Application of the Mobile-Chord Method for the Method of Characteristics International conference

    ICONE15 2007年 

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    Event date: 2007.4

    Language:English   Presentation type:Oral presentation (general)  

  74. Erbia-bearing Super High Burnup Fuel: A Pathway for Breaking 5wt% Enrichment Barrier in LWR Fuel International conference

    ICONE15 2007年 

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    Event date: 2007.4

    Language:English   Presentation type:Oral presentation (general)  

    Country:Japan  

  75. Performance of the Diffusion and Simplified PN Theories For BWR Pin-By-Pin Fine Mesh Core Analyses International conference

    M&C 2007 

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    Event date: 2007.4

    Language:English  

  76. 軽水炉解析における非等方散乱の取り扱い

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    Event date: 2007.3

    Language:Japanese   Presentation type:Oral presentation (general)  

    Country:Japan  

  77. 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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    Event date: 2007

    Language:English  

    Country:Japan  

  78. Application of the Krylov Subspace Method to Burnup Calculation in Lattice Physics Code International conference

    ANS 2006 Winter Meeting 

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    Event date: 2006.11

    Language:English   Presentation type:Oral presentation (general)  

  79. Improvement of the Flat Source Approximation in the Method of Characteristics International conference

    ANS 2006 Winter Meeting 

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    Event date: 2006.11

    Language:English   Presentation type:Oral presentation (general)  

  80. 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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    Event date: 2006.9

    Language:Japanese   Presentation type:Oral presentation (general)  

    Country:Japan  

  81. Generation of Cross Section Library for Lattice Physics Code, AEGIS International conference

    PHYSOR2006 

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    Event date: 2006.9

    Language:English   Presentation type:Oral presentation (general)  

  82. Development of Erbia-bearing Super High Burnup Fuel International conference

    ICAPP06 

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    Event date: 2006.6

    Language:English   Presentation type:Oral presentation (general)  

  83. 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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    Event date: 2006.3

    Language:Japanese   Presentation type:Oral presentation (general)  

    Country:Japan  

  84. Effect of Anisotropic Scattering in PWR/APWR Radial-Reflector Calculations International conference

    ANS 2005 Winter Meeting 

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    Event date: 2005.11

    Language:English   Presentation type:Oral presentation (general)  

  85. Non-equidistant Ray Tracing for the Method of Characteristics International conference

    M&C 2005 

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    Event date: 2005.9

    Language:English   Presentation type:Oral presentation (general)  

  86. Calculation Models of AEGIS, and Advanced Neutronics Solver of Next-Generation International conference

    ANS 2005 Annual Meeting 

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    Event date: 2005.6

    Language:English   Presentation type:Oral presentation (general)  

  87. A New Optimization Algorithm for In-core Fuel Shuffling Sequence of BWR International conference

    ANS 2005 Annual Meeting 

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    Event date: 2005.6

    Language:English   Presentation type:Oral presentation (general)  

  88. 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 

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    Event date: 2005

    Language:Japanese   Presentation type:Oral presentation (general)  

    Country:Japan  

  89. 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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    Event date: 2004

    Language:Japanese   Presentation type:Oral presentation (general)  

    Country:Japan  

  90. 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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    Event date: 2004

    Language:Japanese   Presentation type:Oral presentation (general)  

    Country:Japan  

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KAKENHI (Grants-in-Aid for Scientific Research) 5

  1. 原子力安全に寄与する高速・高精度な次世代燃焼・放射化計算手法の開発

    Grant number:24K08300  2024.4 - 2027.3

    科学研究費助成事業  基盤研究(C)

    山本 章夫

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    Authorship:Principal investigator 

    Grant amount:\4290000 ( Direct Cost: \3300000 、 Indirect Cost:\990000 )

  2. 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

    Grant number:21K04940  2021.4 - 2024.3

    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

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    Authorship:Principal investigator  Grant type:Competitive

    Grant amount:\3900000 ( Direct Cost: \3000000 、 Indirect Cost:\900000 )

  3. Uncertainty quantification of reactor analysis method at design extension condition: A new estimation method based on covariance between input and prediction error

    Grant number:16K06956  2016.4 - 2019.3

    Yamamoto Akio

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    Authorship:Principal investigator 

    Grant amount:\4680000 ( Direct Cost: \3600000 、 Indirect Cost:\1080000 )

    A new evaluation method for prediction error of neutronics simulations under severe accident conditions of a nuclear reactor has been developed. The present method utilizes the Kriging method, which is used in the field of geostatstics, and the principal component analysis. Calculation error of simulation is evaluated by learning the correlation between the parameters used in the simulation and the calculation errors.
    The present method is applied to predict the error of effective multiplication factor of LWR fuel assemblies for various conditions including from normal operation to severe accident conditions. The difference of multiplication factors obtained by the deterministic and the continuous energy Monte-Carlo methods is considered as the calculation error. The results indicate that the present method accurately estimates the calculation error for wide range of reactor conditions.

  4. 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

    2012.4 - 2015.3

    科学研究費補助金  基盤研究(C)

    山本章夫

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    Authorship:Principal investigator 

    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。

  5. 逆解析を用いた燃料配置最適化方法に関する研究

    2007

    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

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    Authorship:Principal investigator 

 

Teaching Experience (On-campus) 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. エネルギー量子制御工学特論

    2016

  19. 基礎セミナーA

    2016

  20. 原子炉物理学

    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. Energy and Environment

    2011

  37. First Year Seminar 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

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