研究者詳細 - 山本　章夫

論文 - 山本　章夫

分割表示 475 件中 81 - 100 件目／全件表示 >>

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 頁： 2219 - 2230 2021年

記述言語：英語掲載種別：研究論文（学術雑誌）出版者・発行元：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

Scopus

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 頁： 1719 - 1728 2021年

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

Scopus

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 頁： 710 - 720 2021年

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

Scopus

Data Assimilation Using Subcritical Measurement of Prompt Neutron Decay Constant

Endo, T; Yamamoto, A

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

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

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

DOI： 10.1080/00295639.2020.1720499

Web of Science

Scopus

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

Foad, B; Yamamoto, A; Endo, T

ANNALS OF NUCLEAR ENERGY 146 巻 2020年10月

記述言語：英語掲載種別：研究論文（学術雑誌）出版者・発行元：Annals of Nuclear Energy

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

DOI： 10.1016/j.anucene.2020.107614

Web of Science

Scopus

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

Foad, B; Yamamoto, A; Endo, T

ANNALS OF NUCLEAR ENERGY 143 巻 2020年8月

記述言語：英語掲載種別：研究論文（学術雑誌）出版者・発行元：Annals of Nuclear Energy

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

DOI： 10.1016/j.anucene.2020.107493

Web of Science

Scopus

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

Matsushita Masaki, Endo Tomohiro, Yamamoto Akio

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

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

Foad, B; Yamamoto, A; Endo, T

ANNALS OF NUCLEAR ENERGY 141 巻 2020年6月

記述言語：英語掲載種別：研究論文（学術雑誌）出版者・発行元：Annals of Nuclear Energy

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

DOI： 10.1016/j.anucene.2020.107341

Web of Science

Scopus

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

Oshima, Y; Endo, T; Yamamoto, A; Kodama, Y; Ohoka, Y; Nagano, H

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

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

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

DOI： 10.1080/00295639.2020.1722512

Web of Science

Scopus

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

Tsujita Kosuke, Endo Tomohiro, Yamamoto Akio

JOURNAL OF NUCLEAR SCIENCE AND TECHNOLOGY 2020年1月

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

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

JOURNAL OF NUCLEAR SCIENCE AND TECHNOLOGY 2020年1月

A kinetic model for the direct response matrix method

Mitsuyasu, T; Yamamoto, A

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

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

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

DOI： 10.1080/00223131.2019.1659874

Web of Science

Scopus

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

Tsujita K., Endo T., Yamamoto A.

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

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

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

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

計算科学を活用した炉物理研究の最先端

山本章夫

年次大会 2020 巻 ( 0 ) 頁： K08101 2020年

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

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

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

記述言語：日本語掲載種別：研究論文（学術雑誌）出版者・発行元：International Conference on Physics of Reactors: Transition to a Scalable Nuclear Future, PHYSOR 2020

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

DOI： 10.1051/epjconf/202124702006

Scopus

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

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

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

記述言語：日本語掲載種別：研究論文（学術雑誌）出版者・発行元：International Conference on Physics of Reactors: Transition to a Scalable Nuclear Future, PHYSOR 2020

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

DOI： 10.1051/epjconf/202124709004

Scopus

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

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

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

記述言語：日本語掲載種別：研究論文（学術雑誌）出版者・発行元：International Conference on Physics of Reactors: Transition to a Scalable Nuclear Future, PHYSOR 2020

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

DOI： 10.1051/epjconf/202124717001

Scopus

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

Yamamoto A., Endo T., Tada K.

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

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

Hayashi T., Endo T., Yamamoto A.

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

　前のページ - 次のページ　

<山本 章夫>

論文 - 山本 章夫

<山本　章夫>

論文 - 山本　章夫