Faculty Profiles - ENDO Tomohiro

MISC - ENDO Tomohiro

Division display 41 - 60 of about 192 ／ All the affair displays >>

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

Masato Ito, Tomohiro Endo, Akio Yamamoto, Yoshitada Masaoka, Yasuhiro Kodama, Hiroaki Nagano

Proceedings of International Conference on Physics of Reactors 2022 (PHYSOR 2022) page： 503 - 512 2022.5

Language：English Publishing type：Research paper, summary (international conference) Publisher：American Nuclear Society

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

Development of Nuclear Data Processing CodeFRENDY Version 2 Reviewed International journal

Kenichi Tada, Akio Yamamoto, Tomohiro Endo, Go Chiba, Michitaka Ono, Masayuki Tojo

Proceedings of International Conference on Physics of Reactors 2022 (PHYSOR 2022) page： 107 - 116 2022.5

Language：English Publishing type：Research paper, summary (international conference) Publisher：American Nuclear Society

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 2including these functions. The present paper explains the overview of FRENDY and features of new functions implemented in FRENDY version 2.

DOI： 10.13182/PHYSOR22-37299

Area Ratio Method Using Dynamic Mode Decomposition Reviewed International journal

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

Proceedings of International Conference on Physics of Reactors 2022 (PHYSOR 2022) page： 3366 - 3375 2022.5

Authorship：Lead author Language：English Publishing type：Research paper, summary (international conference) Publisher：American Nuclear Society

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

Dimension-reduced Nuclear Data Adjustment Method based on the Bayesian Monte-Carlo Method Reviewed International journal

Yuhei Fukui, Tomohiro Endo, Akio Yamamoto

Transactions of the American Nuclear Society Vol. 125 ( 1 ) page： 900 - 903 2021.11

Application of Regionwise Even-Parity Discontinuity Factor to the Multigroup Analog Monte-Carlo Method Reviewed International journal

Yoshiki Oshima, Tomohiro Endo, Akio Yamamoto, Naoto Aizawa

Transactions of the American Nuclear Society Vol. 125 ( 1 ) page： 904 - 907 2021.11

Sensitivity Analysis in Dynamic Multi-unit PRA Evaluation Using Correlated Sampling Method Reviewed International journal

Yuki Morishita, Akio Yamamoto, Tomohiro Endo

Proc. Asian Symposium on Risk Assessment and Management (ASRAM2021) page： ASRAM2021-034 2021.10

Language：English Publishing type：Research paper, summary (international conference)

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 failure rate of safety components. In the correlated sampling method, original and perturbed samples are assumed to trace the identical accident sequence, but the weight of the perturbed sample is adjusted to incorporate the variation of input data. The correlated sampling method is applied to the sensitivity analysis of a simple system to prove the principle. The result indicates that the sensitivity analysis for the CMMC coupling method is possible by the correlated sampling method.

Development of estimation method for prompt neutron decay constant using dynamic mode decomposition Reviewed International journal

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

Proc. M&C 2021 page： 1719 - 1728 2021.10

Language：English Publishing type：Research paper, summary (international conference) Publisher：American Nuclear Society

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

Theoretical derivation of unique combination-number for higher order neutron correlation factors based on Pál-Bell equation Invited Reviewed International journal

Tomohiro Endo, Akio Yamamoto

Proc. M&C 2021 page： 1568 - 1576 2021.10

Authorship：Lead author,　Corresponding author Language：English Publishing type：Research paper, summary (international conference) Publisher：American Nuclear Society

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

S2 Consistent Analytic CMFD Acceleration for Method of Characteristics Reviewed International journal

Yoshiki Oshima, Akio Yamamoto, Tomohiro Endo

Proc. M&C 2021 page： 1840 - 1849 2021.10

Language：English Publishing type：Research paper, summary (international conference) Publisher：American Nuclear Society

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

Fuel assembly analyses with resonance calculation using energy spectrum expansion method Reviewed International journal

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

Proc. M&C 2021 page： 2219 - 2230 2021.10

Language：English Publishing type：Research paper, summary (international conference) Publisher：American Nuclear Society

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

FRENDY/MG: a multi-group cross section generation module using ace pointwise cross sections Reviewed International journal

Akio Yamamoto, Tomohiro Endo, Basma Foad, Go Chiba, Kenichi Tada

Proc. M&C 2021 page： 710 - 720 2021.10

Language：English Publishing type：Research paper, summary (international conference) Publisher：American Nuclear Society

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

Multi-Group Cross Section Library Generation by FRENDY for Fast Reactor Neutronics Calculations Reviewed International journal

Go Chiba, Akio Yamamoto, Kenichi Tada, Tomohiro Endo

Transactions of the American Nuclear Society Vol. 124 page： 556 - 558 2021.6

Verification of the Multi-Group Generation Capability of FRENDY Nuclear Data Processing Code for Recent Nuclear Data through Comparison of One-group Reaction Rates Reviewed International journal

Akio Yamamoto, Kenichi Tada, Go Chiba, Tomohiro Endo

Transactions of the American Nuclear Society Vol. 124 page： 544 - 547 2021.6

Perturbation-Theory-Based Sensitivity Analysis of Prompt Neutron Decay Constant for Water-Only System Reviewed International journal

Tomohiro Endo, Akihiro Noguchi, Akio Yamamoto, Kenichi Tada

Transactions of the American Nuclear Society Vol. 124 page： 184 - 187 2021.6

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 Reviewed International journal

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

EPJ Web of Conferences Vol. 247 page： 02006 - 02006 2021.2

Language：English Publishing type：Research paper, summary (international conference) Publisher：EDP Sciences

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

Experiment of Unique Combination Number due to the Third-Order Neutron-Correlation Reviewed International journal

Tomohiro Endo, Sho Imai, Kenichi Watanabe, Akio Yamamoto, Atsushi Sakon, Kengo Hashimoto, Masao Yamanaka, Cheol Ho Pyeon

EPJ Web of Conferences Vol. 247 page： 09004 - 09004 2021.2

Authorship：Lead author,　Corresponding author Language：English Publishing type：Research paper, summary (international conference) Publisher：EDP Sciences

From zero-power reactor noise measurement, the second- and third-order neutron correlation factors <italic>Y</italic> and <italic>y</italic>3 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 <italic>y</italic>3/<italic>Y</italic>2 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 <italic>T</italic> increases sufficiently. Thus, the information about the difference between <italic>y</italic>3/<italic>Y</italic>2 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 <italic>y</italic>3/<italic>Y</italic>2 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 <italic>y</italic>3/<italic>Y</italic>2 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 <italic>y</italic>3/<italic>Y</italic>2 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 <italic>y</italic>3/<italic>Y</italic>2 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

Estimated Criticality Lower-Limit Multiplication Factor of Low-Enriched Uranium Dioxide-Concrete System Using the Bootstrap Method Reviewed International journal

Takuto Hayashi, Fuga Nishioka, Tomohiro Endo, Akio Yamamoto

EPJ Web of Conferences Vol. 247 page： 17001 - 17001 2021.2

Language：English Publishing type：Research paper, summary (international conference) Publisher：EDP Sciences

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

Basic Study on a Pseudo Trend Phenomenon of The Feynman-α Analysis with Bunching Method Reviewed International journal

Sin-ya Hohara, Atsushi Sakon, Tomohiro Endo, Tadafumi Sano, Kunihiro Nakajima, Kazuki Takahashi, Kengo Hashimoto

EPJ Web of Conferences Vol. 247 page： 09008 - 09008 2021.2

Language：English Publishing type：Research paper, summary (international conference) Publisher：EDP Sciences

In these years, reactor noise analysis methods have been studied to apply for the Debris’ criticality management at the Fukushima Daiichi NPP, Japan. The Feynman-α analysis with bunching method is one of the candidate techniques, however the bunching method itself has never been validated in detail. This synthesis technique is useful to reduce a time required for the experiment, however it is known that a non-physical trend unrelated to the state of a nuclear reactor is generated by the multiple use of time series data, and this phenomenon (we call “pseudo trend phenomenon”) has never been systematically studied in detail. In this study, Poisson events, whose statistical characteristics were clarified, were employed to investigate the pseudo trend phenomenon of the bunching method. The time-sequence count data for various statistical parameters were generated by the Monte Carlo time series simulator. Comparing the two results obtained by applying the conventional bunching method and the moving-bunching method for the same Poisson event time series, and it was found that the same pseudo trend component appears in both results of the bunching method and the moving bunching method. In addition, it was also found that the fluctuation width of the pseudo trend component is smaller than the statistical fluctuation range.

DOI： 10.1051/epjconf/202124709008

Study on Multi-unit PRA Considering Correlation between Multiple Safety Systems Using the Continuous Markov chain Monte Carlo Method Reviewed International journal

Kento Sawada, Tomohiro Endo, Akio Yamamoto, Sunghyon Jang, Daisuke Fujiwara, Chikahiro Sato

Proc. ASRAM2020 page： 1027 2020.11

Language：English Publishing type：Research paper, summary (international conference) Publisher：Atomic Energy Society of Japan

After the accident in Fukushima Dai-ichi Nuclear Power Plants, the necessity for the safety assessment considering multiple nuclear power units is increasing. However, the conventional event tree (ET) method, which is widely used in probabilistic risk assessment (PRA), cannot consider the interdependency among safety systems or events in multiple units as well as time dependency of the accident progressions. Recently, a new approach that couples the Continuous Markov chain Monte Carlo (CMMC) method with a plant safety analysis or a severe accident analysis code is suggested to address these issues. However, applications of the CMMC coupling method to safety assessment for multiple units are still limited. In the present study, the CMMC coupling method is applied to multiple units analysis in order to get practical insights and to clarify the issues of this method. Since the CMMC coupling method requires a large number of executions of an accident analysis code, a meta-model that simplifies the systems and the physical phenomena in accidents is used in this study to avoid the high computational cost of the existing accident analysis codes. In the present study, the numerical investigation of a BWR accident under station blackout (SBO) with loss of RCIC is carried out. The preliminary results suggested that the CMMC coupling method can appropriately treat the interdependency and time dependency among events in multiple units.

Subcriticality Estimation using Unscented Kalman Filter for Reactivity- and Source-Transients Reviewed International journal

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

Transactions of the American Nuclear Society Vol. 123 page： 841 - 844 2020.11

　 PREV - NEXT 　

<ENDO Tomohiro>

MISC - ENDO Tomohiro