研究者詳細 - 遠藤　知弘

MISC - 遠藤　知弘

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遠藤知弘, 左近敦士

アトモス : 日本原子力学会誌61 巻 ( 12 ) 頁： 857 - 862 2019年12月

担当区分：筆頭著者,　責任著者記述言語：日本語掲載種別：記事・総説・解説・論説等（商業誌、新聞、ウェブメディア）出版者・発行元：日本原子力学会

原子炉物理学分野において，これまでに提案されてきた未臨界度測定手法について解説し，既存手法の課題を述べる。未臨界度測定法は，静的手法，動的手法，炉雑音解析手法に分類される。各手法にはそれぞれ利点・欠点があるため，対象とする測定体系や状況に応じて適切な測定手法を選定したり，組み合わせて使用したりすることが重要である。

DOI： 10.3327/jaesjb.61.12_857

CiNii Books

2D Cylindrical SP3 Transport Calculation Using Finite Volume Method with Cross-Diffusion Treatment for Unstructured Meshes 査読有り国際誌

Tomohiro Endo, Akio Yamamoto

Transactions of the American Nuclear Society131 巻 ( 1 ) 頁： 1053 - 1056 2024年11月

担当区分：筆頭著者記述言語：英語掲載種別：研究発表ペーパー・要旨（国際会議）出版者・発行元：American Nuclear Society

This study presents a prototype of an in-house code to numerically solve the simplified P3 (SP3) neutron transport equation for a 2D cylindrical (r-z) geometry using unstructured meshes for spatial discretization. In the developed SP3 calculation code, the cross-diffusion effect due to the unstructured meshes can be treated by a variant of the Green-Gauss reconstruction technique.

DOI： 10.13182/T131-46095

Study on Fusion Safety Concept Based on Nuclear Safety (2) Defense in Depth in Fusion Safety 査読有り国際誌

Yoshiki Yamashita, Tomohiro Endo, Akio Yamamoto, Takashi Ino

Transactions of the American Nuclear Society131 巻 ( 1 ) 頁： 299 - 302 2024年11月

記述言語：英語掲載種別：研究発表ペーパー・要旨（国際会議）出版者・発行元：American Nuclear Society

The concept of fusion safety is studied with reference to nuclear safety. Risk profiles of fusion reactor facilities are reviewed and organized and implementation of defense in
depth (DID) in fusion safety is developed based on the risk profiles.

DOI： 10.13182/T131-46140

Study on Fusion Safety Concept Based on Nuclear Safety (1) Fusion Safety Objective and Principles 査読有り国際誌

Yoshiki Yamashita, Tomohiro Endo, Akio Yamamoto, Takashi Ino

Transactions of the American Nuclear Society131 巻 ( 1 ) 頁： 292 - 295 2024年11月

A Simple Resonance Calculation Approach for Space-Dependent Self-Shielding 査読有り国際誌

Akio Yamamoto, Tomohiro Endo

Transactions of the American Nuclear Society131 巻 ( 1 ) 頁： 1038 - 1041 2024年11月

記述言語：英語掲載種別：研究発表ペーパー・要旨（国際会議）出版者・発行元：American Nuclear Society

A simple resonance calculation method for space-dependent self-shielding treatment is proposed. The present method is based on three calculations: 1) ultra-fine group (UFG) calculation with the narrow resonance (NR) approximation in a homogeneous medium, 2) UFG calculation with explicit slowing down (SLD) calculation in a homogeneous medium, and 3) UFG calculation with the NR approximation in a heterogeneous system. The space-dependent effective cross sections are generated by the synthesis of these three calculations.

DOI： 10.13182/T131-45885

Implementation of track length estimator for flux distribution tallies using proper orthogonal decomposition in one-dimensional geometry 査読有り国際誌

Ryoichi Kondo, Tomohiro Endo, Akio Yamamoto

EPJ Web of Conferences302 巻頁： 04002 2024年10月

記述言語：英語掲載種別：研究発表ペーパー・要旨（国際会議）出版者・発行元：EDP Sciences

In a recent study, we have developed an efficient flux distribution tallying method in the Monte Carlo calculation toward the high-fidelity, large scale multi-physics simulation. In this method, the proper orthogonal decomposition is applied to the flux distribution tallies. While the tallying method was implemented with the collision estimator in the previous study, the track length estimator is implemented in the present study to obtain the tally with lower statistical error. The implementation of the flux distribution tally with the track length estimator is compared with that of the collision estimator and the normal track length estimator in a one-dimensional problem. The numerical results reveal that the distribution tally using the proper orthogonal decomposition with the track length estimator can obtain a more precise solution compared with that with the collision estimator. Therefore, in terms of the statistical error, the relationship between the distribution tally with track length and collision estimators is similar to that between the conventional track length and collision estimators.

DOI： 10.1051/epjconf/202430204002

Data Assimilation Using Deterministic Sampling Method to Selectively Reduce Uncertainty due to Thermal Neutron Scattering Law for Light Water 査読有り国際誌

Yoshinari Harada, Hibiki Yamaguchi, Tomohiro Endo, Akio Yamamoto, Kenichi Tada

Transactions of the American Nuclear Society130 巻 ( 1 ) 頁： 758 - 762 2024年6月

記述言語：英語掲載種別：研究発表ペーパー・要旨（国際会議）出版者・発行元：American Nuclear Society

This study presents data assimilation using the prompt neutron decay constant α of the water tank system to selectively reduce the uncertainties due to the thermal neutron scattering law of light water. For efficient uncertainty quantification and data assimilation, the deterministic sampling method was applied in this study. Furthermore, in the data assimilation, the α measurement of the water tank system was utilized to improve the prediction accuracy of keff calculations in a critical experiment involving light water.

DOI： 10.13182/T130-44516

Loading Pattern Optimization with Deep Learning and Monte-Carlo Tree Search 査読有り国際誌

Rikuto Kasama, Akio Yamamoto, Tomohiro Endo

Transactions of the American Nuclear Society130 巻 ( 1 ) 頁： 1044 - 1047 2024年6月

記述言語：英語掲載種別：研究発表ペーパー・要旨（国際会議）出版者・発行元：American Nuclear Society

We have developed a new optimization method based on Deep Learning (DL) and Monte-Carlo Tree Search (MCTS) for fuel loading pattern optimization. The fuel loading pattern optimization is redefined as a sequence of fuel loadings to an empty core, and the sequence is optimized by the DL and MCTS approach, which is used in AI for board games such as Go, Chess, and Shogi (Japanese Chess).

DOI： 10.13182/T130-44319

Statistical Error Estimation of Autocorrelation Method Using Circular Block Bootstrap Method 査読有り国際誌

Ryoga Hirota, Tomohiro Endo, Akio Yamamoto, Kenichi Watanabe, Junichi H. Kaneko

Proc. PHYSOR2024 頁： 78 - 86 2024年4月

記述言語：英語掲載種別：研究発表ペーパー・要旨（国際会議）出版者・発行元：American Nuclear Society

The autocorrelation method is a subcriticality measurement technique based on the reactor noise analysis method. In this method, the prompt neutron decay constant can be obtained from the exponential decay of the autocorrelation of successively detected neutron counts in a target subcritical system. To simply estimate the statistical error of the prompt neutron decay constant, several measurements of the reactor noise are required, although the total measurement time is inevitably long for typical systems where the reactor noise analysis method is applied. Therefore, the purpose of this study is to investigate the applicability of the circular block bootstrap method in order to estimate the statistical error of the prompt neutron decay constant obtained by the autocorrelation method using a single reactor noise measurement. The bootstrap-based statistical error estimation method is validated using the time series data of reactor noise measurements at UTR-KINKI in a shutdown state with the inherent neutron source in the uranium-aluminum fuel. Consequently, this study demonstrates that the circular block bootstrap method can be also utilized for the autocorrelation method, to estimate the confidence interval of the prompt neutron decay constant as the statistical error. Namely, a single reactor noise measurement can be effectively reused for error estimation instead of multiple measurements.

その他リンク： https://www.ans.org/meetings/physor2024/sessions/attachment/paper-9424/

Investigation of Subcriticality Monitoring Method Using Improved Simplest Reactivity Estimator with Bilateral Filter 査読有り国際誌

Taiyo Moribe, Tomohiro Endo, Akio Yamamoto, Junichi H. Kaneko

Proc. PHYSOR2024 頁： 97 - 106 2024年4月

記述言語：英語掲載種別：研究発表ペーパー・要旨（国際会議）出版者・発行元：American Nuclear Society

In the retrieval of fuel debris from the Fukushima Daiichi nuclear power plant, there are the following issues: lack of information needed to inversely estimate the subcriticality; and the limitation of neutron detector size. In order to address these issues, we propose a reactivity estimation method that combines the improved simplest reactivity estimator (SRE) and the least squares inverse kinetics method. This study aims to estimate the reactivity in dollar units in a data driven manner only from the measured neutron count rate without point kinetics parameters. Furthermore, the usefulness of the bilateral filter in combination with the median for the count rate is also investigated to mitigate statistical errors in the reactivity estimation results due to the statistical fluctuation and outlier in the measured count rate data. To verify our proposed method, point kinetics calculation is performed to simulate the time series data of the neutron count rate during a stepwise transient from a deep subcritical state to the shallower state. Here, the magnitude of the count rate at the initial state is parametrically varied to investigate the effect of the statistical count rate fluctuation on the estimated reactivity. As a result, it is observed that the estimated reactivity tends to be overestimated (i.e., closer to critical) as lower count rate data are used in our proposed method. Thus, our proposed method is considered practical from the perspective of criticality control. Moreover, it is clarified that the combinational use of the bilateral filter and the median can effectively reduce the error of the estimated reactivity due to the statistical fluctuation of the count rate. Consequently, this feasibility study demonstrates our proposed method for monitoring the reactivity change even under the low neutron count rate condition.

DOI： 10.13182/PHYSOR24-43487

Loading Pattern Optimization for LWRs Using Monte Carlo Tree Search 査読有り国際誌

Rikuto Kasama, Akio Yamamoto, Tomohiro Endo

Proc. PHYSOR2024 頁： 2490 - 2499 2024年4月

記述言語：英語掲載種別：研究発表ペーパー・要旨（国際会議）出版者・発行元：American Nuclear Society

Inspired by recent breakthroughs in AI for games, we transform the fuel loading pattern
optimization problem into a game tree search problem and adopt Monte Carlo Tree Search (MCTS) to solve it. The performance of MCTS is compared with the conventional optimization methods, such as the random search and simulated annealing (SA), for the loading pattern optimization of a typical pressurized water reactor. The results demonstrate the superior efficiency of MCTS over the conventional methods. MCTS consistently produces high-quality patterns and effectively avoids the common problem of falling into local minima. By using techniques rooted in AI, we are
opening up a new way of solving complex problems in nuclear engineering, offering the prospect of more reliable and efficient core designs. This innovative approach will not only benefit the nuclear field, but also serve as a bridge between optimization and artificial intelligence.

DOI： 10.13182/PHYSOR24-43466

Preliminary Study on Two-Dimensional SP3 Calculation Based on POD-Local/Global Iterations 査読有り国際誌

Masato Ito, Akio Yamamoto, Tomohiro Endo, Yasuhiro Kodama, Taichi Takeishi, Hiroaki Nagano

Proc. PHYSOR2024 頁： 1612 - 1621 2024年4月

担当区分：責任著者記述言語：英語掲載種別：研究発表ペーパー・要旨（国際会議）出版者・発行元：American Nuclear Society

As a preliminary study for an efficient heterogeneous neutron transport calculation, this study newly proposes a reduced order core calculation method by applying proper orthogonal decomposition (POD) to the SP3 calculation with the global/local iterations. The fine mesh SP3 calculation (local calculation) for each single assembly can be efficiently solved using POD, thanks to the matrix form-based SP3 equation and the relatively small scale of the simultaneous equations. By reconstructing the fine mesh distribution of neutron flux and partial neutron currents from the POD bases and the expansion coefficients estimated by the local calculation, the p-CMFD calculation is applied to the coarse mesh calculation (global calculation). These local and global calculations are iterated until convergence. Through the two-dimensional small and large UO2-MOX core calculations, our proposed method is verified by comparing with the reference result by the conventional fine mesh SP3 calculation.

DOI： 10.13182/PHYSOR24-43629

Real-Time 3D Fine Spatial Mesh Kinetics Simulator Using POD for Coupled Core 査読有り国際誌

Kaito Ito, Kosuke Tsujita, Tomohiro Endo, Akio Yamamoto

Proc. PHYSOR2024 頁： 6986 2024年4月

担当区分：責任著者記述言語：英語掲載種別：研究発表ペーパー・要旨（国際会議）出版者・発行元：American Nuclear Society

Toward the development of the digital triplets to effectively educate the reactor physics experiment, the present study develops a real-time 3D fine spatial mesh kinetics simulator based on the proper orthogonal decomposition (POD). Our developed simulator, called Ikaros3D, was specifically designed for a coupled core to promote students’ better understanding of the complicated behavior of the spatial variation of the dynamic reactivity depending on the neutron detector position. Through a verification test, we confirmed that the reduced order model (ROM) using POD was able to accurately and quickly simulate the variation in the total core power due to the operation of two different control rods, thanks to the pre-tabulated compressed coefficient matrices for the POD based kinetics calculation.

DOI： 10.13182/PHYSOR24-43834

Reconstruction of In-Core Power Distribution Based on POD Using Ex-Core Detector Signals 査読有り国際誌

Yuki Urase, Tomohiro Endo, Akio Yamamoto

Proc. PHYSOR2024 頁： 1642 - 1651 2024年4月

記述言語：英語掲載種別：研究発表ペーパー・要旨（国際会議）出版者・発行元：American Nuclear Society

A reconstruction method of in-core power distribution using the proper orthogonal decomposition (POD) and the signal from ex-core detectors is developed. In the present method, the neutron flux distribution is expanded in a small number of POD bases and expansion coefficients. Neutron transport calculations were performed with several calculation conditions to perturb the neutron flux distributions in the reactor core and the POD bases were obtained using the multiple neutron flux distributions as the snapshot data. The neutron counts of ex-core detectors were evaluated from the in-core neutron flux distributions and the detector response coefficients. Considering the neutron counts of ex-core detectors as measured signals, the expansion coefficients of POD that represent the in-core neutron flux distributions were determined. Assuming a simple two-dimensional core with nine regions, the reconstruction accuracy of in-core neutron flux distribution was verified for several conditions. The results indicate that the in-core neutron flux distributions can be reconstructed by the ex-core detector signals when the distance between the detectors and the core is appropriately chosen.

DOI： 10.13182/PHYSOR24-43478

Simplified Treatment of Coating Layers in TRISO Fuel in Statistical Geometry Method in Monte Carlo Calculation 査読有り国際誌

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

Proc. PHYSOR2024 頁： 1036 - 1047 2024年4月

記述言語：英語掲載種別：研究発表ペーパー・要旨（国際会議）出版者・発行元：American Nuclear Society

An improved sampling method for flight distance is proposed for Monte Carlo analysis of TRISO fuel particles using the statistical geometry method. The statistically uniform distribution of fuel particles, which is usually assumed as a default sampling method of flight distance of a neutron in graphite matrix, shows considerable bias on neutronics property when coating layers of a TRISO fuel particle are homogenized with graphite matrix. The proposed new sampling method almost resolves the difference between the homogenized coating model and the reference model that explicitly considers coating layers. By adopting the present method, the homogenized coating model can be used without significant loss of accuracy in the statistical geometry method. Computation time for a typical HTTR fuel compact cell with a continuous energy Monte Carlo code is reduced to 1/7 when the homogenized coating model is used.

DOI： 10.13182/PHYSOR24-43318

α-eigenvalue Calculation using the Sn Method with GPU Diffusion Acceleration 査読有り国際誌

H. Yamaguchi, T. Endo, A. Yamamoto

Proceedings of the Reactor Physics Asia 2023 (RPHA2023) Conference 2023年10月

2D Neutron Diffusion Calculation Based on Local/Global Iterations Using Proper Orthogonal Decomposition 査読有り国際誌

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

Proceedings of the Reactor Physics Asia 2023 (RPHA2023) Conference 2023年10月

Feasibility Study on Subcritical Rod Worth Measurement for UTR-KINKI 査読有り国際誌

T. Endo, G. Chiba, K. Watanabe, C. H. Pyeon, G. Wakabayashi

Proceedings of the Reactor Physics Asia 2023 (RPHA2023) Conference 2023年10月

担当区分：筆頭著者,　責任著者記述言語：英語掲載種別：研究発表ペーパー・要旨（国際会議）出版者・発行元：Korean Nuclear Society

In this study, we investigate the feasibility of the subcritical rod worth measurement (SRWM) to UTRKINKI (University Training Reactor, Kindai University) through a virtual numerical experiment.

Comparison of Neutronic Characteristics of BWR Burnup Fuel between JENDL-4.0 and JENDL-5 査読有り国際誌

Tomoaki Watanabe, Kenichi Tada, Tomohiro Endo, Akio Yamamoto

Proc. ICNC2023 2023年10月

記述言語：英語掲載種別：研究発表ペーパー・要旨（国際会議）出版者・発行元：Japan Atomic Energy Agency

The present study investigated the effect of the nuclear data modification from JENDL-4.0 to JENDL-5 on the criticality characteristics of a typical 9×9 BWR burnup fuel. In JENDL-5, many nuclear data essential to characterize burnup fuels, such as the cross sections of Gd-155 and Gd-157, were modified. The depletion calculations showed that kinf of JENDL-5 was smaller than that of JENDL-4.0 at almost all burnup steps for 0–50 GWd/t. Significant burnup dependencies of the kinf difference between JENDL-4.0 and JENDL-5 were also found. For example, the kinf difference significantly increased around 12 GWd/t, and the kinf difference gradually increased as the burnup value increased. The direct sensitivity calculations were carried out to specify which nuclear data impacts the kinf difference. The specific nuclear data of interest was replaced from JENDL-4.0 to JENDL-5. The sensitivity calculation results showed that the kinf difference around 12 GWd/t was mainly caused by the modification of the cross sections of U-235, Gd-155, Gd-157, and TSL data of H in H2O. These modifications affected the composition of Gd-155 and Gd-157. The sensitivity calculation results also clarified that modifying the U-235 and Pu-239 cross sections and TSL data of H in H2O affected the increase in the kinf difference as the burnup value increased. The modifications of the TSL data for H in H2O also affected the composition of U-235.

Program of Reactor Physics Experiment to Measure Subcriticality for an Unknown System 査読有り国際誌

Shunya Teratani, Yoshinari Harada, Kaito Mori, Yoshiki Yamashita, Tomohiro Endo, Go Chiba

Proc. ICNC2023 2023年10月

記述言語：英語掲載種別：研究発表ペーパー・要旨（国際会議）出版者・発行元：Japan Atomic Energy Agency

In order to train human resources for the future nuclear field, we have been developing an advanced educational program for the reactor physics experiments by utilizing a Japanese educational reactor (University Training Reactor, Kindai University, UTR-KINKI) with recent measurement techniques. This paper summarizes the advanced reactor physics experiment program for training graduate students to tackle unknown problems. As a problem-based learning program (PBL) of reactor physics experiments, we focused on the subcriticality measurement for an unknown system without prior knowledge of the external neutron source intensity and the point kinetics parameters. In order to solve this challenging problem, the basic lectures and the interactive exercise were prepared to help graduate students better understand (1) the criticality approach based on the inverse multiplication and (2) the measurement of the prompt neutron decay constant by the Feynman-α method. Through the series of lectures, the instructors hope that graduate students will get an idea of how to determine the subcriticality for a target unknown system, e.g., by combining the inverse multiplication and the Feynman-α methods through the Simmons-King method. The combined method was demonstrated by the actual reactor physics experiment at UTR-KINKI.

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