Faculty Profiles - YAMAMOTO, Akio

Papers - YAMAMOTO, Akio

Division display 61 - 80 of about 481 ／ All the affair displays >>

山本章夫

日本原子力学会誌ATOMOΣ Vol. 64 ( 3 ) page： 176 - 176 2022

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

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

Language：English Publishing type：Research paper (scientific journal) Publisher：Proceedings of the International Conference on Physics of Reactors, PHYSOR 2022

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

DOI： 10.13182/PHYSOR22-37557

Scopus

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

Language：English Publishing type：Research paper (scientific journal) Publisher：Proceedings of the International Conference on Physics of Reactors, PHYSOR 2022

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

DOI： 10.13182/PHYSOR22-37693

Scopus

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

Language：English Publishing type：Research paper (scientific journal) Publisher：Proceedings of the International Conference on Physics of Reactors, PHYSOR 2022

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

DOI： 10.13182/PHYSOR22-37287

Scopus

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

Language：English Publishing type：Research paper (scientific journal) Publisher：Proceedings of the International Conference on Physics of Reactors, PHYSOR 2022

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

DOI： 10.13182/PHYSOR22-37299

Scopus

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

Language：English Publishing type：Research paper (scientific journal) Publisher：Proceedings of the International Conference on Physics of Reactors, PHYSOR 2022

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

DOI： 10.13182/PHYSOR22-37184

Scopus

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

Language：English Publishing type：Research paper (scientific journal) Publisher：Proceedings of the International Conference on Physics of Reactors, PHYSOR 2022

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

DOI： 10.13182/PHYSOR22-37251

Scopus

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 2021.12

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

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

DOI： 10.1080/00223131.2021.1944930

Web of Science

Scopus

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 2021.12

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

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

DOI： 10.1080/00223131.2021.1940341

Web of Science

Scopus

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 2021.11

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

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

DOI： 10.1080/00223131.2021.1921631

Web of Science

Scopus

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 2021.9

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

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

DOI： 10.1080/00223131.2021.1902416

Web of Science

Scopus

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 2021.7

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

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

DOI： 10.1080/00295639.2020.1863066

Web of Science

Scopus

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 2021.2

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

Tran, VP; Phan, GTT; Hoang, VK; Ha, PNV; Yamamoto, A; Tran, HN

ANNALS OF NUCLEAR ENERGY Vol. 151 2021.2

Language：English Publishing type：Research paper (scientific journal) Publisher：Annals of Nuclear Energy

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

DOI： 10.1016/j.anucene.2020.107938

Web of Science

Scopus

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

Sasaki, M; Sanada, Y; Katengeza, EW; Yamamoto, A

SCIENTIFIC REPORTS Vol. 11 ( 1 ) 2021.1

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

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

DOI： 10.1038/s41598-021-81546-4

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 2021.1