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Journal Articles

Report on the 35th Meeting of Working Party on International Nuclear Data Evaluation Co-operation (WPEC)

Iwamoto, Osamu; Iwamoto, Nobuyuki; Kimura, Atsushi; Tada, Kenichi; Yokoyama, Kenji

Kaku Deta Nyusu (Internet), (136), 6 Pages, 2023/10

no abstracts in English

Journal Articles

JSME series in thermal and nuclear power generation Vol.3 (Sodium-cooled fast reactor development; R&Ds on thermal-hydraulics and safety assessment towards social implementation)

Tanaka, Masaaki; Uchibori, Akihiro; Okano, Yasushi; Yokoyama, Kenji; Uwaba, Tomoyuki; Enuma, Yasuhiro; Wakai, Takashi; Asayama, Tai

Dai-27-Kai Doryoku, Enerugi Gijutsu Shimpojiumu Koen Rombunshu (Internet), 5 Pages, 2023/09

The book, JSME Series in Thermal and Nuclear Power Generation Vol.3 Sodium-cooled Fast Reactor, was published as a 30th anniversary memorial project of Power & Energy Systems Division. This paper describes an introduction of the book on a part of key technologies regarding safety assessment, thermal-hydraulics, neutronics, and fuel and material development. This introductory paper also provides an overview of an integrated evaluation system named ARKADIA to offer the best possible solutions for challenges arising during the design process, safety assessment, and operation of a nuclear plant over its life cycle, in active use of the R&D efforts and knowledges on thermal-hydraulics and safety assessment with state-of-the-art numerical analysis technologies.

Journal Articles

Validation practices of multi-physics core performance analysis in an advanced reactor design study

Doda, Norihiro; Kato, Shinya; Hamase, Erina; Kuwagaki, Kazuki; Kikuchi, Norihiro; Ohgama, Kazuya; Yoshimura, Kazuo; Yoshikawa, Ryuji; Yokoyama, Kenji; Uwaba, Tomoyuki; et al.

Proceedings of 20th International Topical Meeting on Nuclear Reactor Thermal Hydraulics (NURETH-20) (Internet), p.946 - 959, 2023/08

An innovative design system named ARKADIA is being developed to realize the design of advanced nuclear reactors as safe, economical, and sustainable carbon-free energy sources. This paper focuses on ARKADIA-Design for design studies as a part of ARKADIA and introduces representative verification methods for numerical analysis methods of the core design. ARKADIA-Design performs core performance analysis of sodium-cooled fast reactors using a multiphysics approach that combines neutronics, thermal-hydraulics, core mechanics, and fuel pin behavior analysis codes. To confirm the validity of these analysis codes, validation matrices are identified with reference to experimental data and reliable numerical analysis results. The analysis models in these codes and the representative practices for the validation matrices are described.

JAEA Reports

Differential pressure rise event for filters of HTTR primary helium gas circulators, 1; Investigation of differential pressure rise event

Nemoto, Takahiro; Arakawa, Ryoki; Kawakami, Satoru; Nagasumi, Satoru; Yokoyama, Keisuke; Watanabe, Masashi; Onishi, Takashi; Kawamoto, Taiki; Furusawa, Takayuki; Inoi, Hiroyuki; et al.

JAEA-Technology 2023-005, 33 Pages, 2023/05


During shut down of the HTTR (High Temperature engineering Test Reactor) RS-14 cycle, an increasing trend of filter differential pressure for the helium gas circulator was observed. In order to investigate this phenomenon, the blower of the primary helium purification system was disassembled and inspected. As a result, it is clear that the silicon oil mist entered into the primary coolant due to the deterioration of the charcoal filter performance. The replacement and further investigation of the filter are planning to prevent the reoccurrence of the same phenomenon in the future.

Journal Articles

General-purpose nuclear data library JENDL-5 and to the next

Iwamoto, Osamu; Iwamoto, Nobuyuki; Kunieda, Satoshi; Minato, Futoshi; Nakayama, Shinsuke; Kimura, Atsushi; Nakamura, Shoji; Endo, Shunsuke; Nagaya, Yasunobu; Tada, Kenichi; et al.

EPJ Web of Conferences, 284, p.14001_1 - 14001_7, 2023/05

 Times Cited Count:0 Percentile:0.21(Nuclear Science & Technology)

Journal Articles

Development of element functions and design optimization procedures with knowledge- and AI-aided design integration approach for advanced reactor lifecycle in ARKADIA

Tanaka, Masaaki; Enuma, Yasuhiro; Okano, Yasushi; Uchibori, Akihiro; Yokoyama, Kenji; Seki, Akiyuki; Wakai, Takashi; Asayama, Tai

Proceedings of 30th International Conference on Nuclear Engineering (ICONE30) (Internet), 11 Pages, 2023/05

Journal Articles

Applicability study of Bayesian optimization to neutronic design of a homogeneous two-region core

Kuwagaki, Kazuki; Yokoyama, Kenji

Proceedings of 30th International Conference on Nuclear Engineering (ICONE30) (Internet), 9 Pages, 2023/05

At the Japan Atomic Energy Agency (JAEA), a design support tool for advanced nuclear reactors is currently under development. This tool is called ARKADIA-Design, and is expected to support the integrated design evaluation of reactors from the viewpoints of safety, economy, and sustainability as a carbon-free energy source by utilizing the newest analysis/evaluation technologies such as AI technology, and the accumulated knowledge of fast reactor development. One development task of the ARKADIA-Design is to build a system that automatically identifies optimized design parameters by which an objective function specified by core performance is minimized (or maximized). In the present study, we set up a single objective optimization example problem with multiple constraints for a homogeneous two-region core, and showed that the optimal solution of this example problem can be automatically obtained by the Bayesian optimization method, which is a candidate optimization algorithm for the system. In addition, we also demonstrated how the system would assist the core design procedure in future, by indirectly solving a three-variable optimization problem of the core design. From these results and demonstrations, we confirmed that the system to be developed has the potential to be a useful support tool for the designers, enabling them to obtain optimal core designs efficiently.

Journal Articles

Investigation of optimization process for core design with integrated analysis between neutronics and plant dynamics

Hamase, Erina; Kuwagaki, Kazuki; Doda, Norihiro; Yokoyama, Kenji; Tanaka, Masaaki

Proceedings of 30th International Conference on Nuclear Engineering (ICONE30) (Internet), 10 Pages, 2023/05

To innovate a core design process, an optimization process for the core design has been developed as a part of the design optimization support tool named ARKADIA-Design. The core design optimization process is integrated by the core design analysis of neutronics, thermal-hydraulics, and fuel integrity and plant dynamics analysis with the Bayesian optimization (BO) algorithm. The optimization problem for design parameters with high core performance and inherent safety in ULOF event was solved by the integrated analysis between the neutronics and plant dynamics with the BO in a primary loop system including a core consisting of two-dimensional RZ cylindrical geometry. It was indicated that the optimization process could obtain an optimal solution.

Journal Articles

JENDL-5 benchmarking for fission reactor applications

Tada, Kenichi; Nagaya, Yasunobu; Taninaka, Hiroshi; Yokoyama, Kenji; Okita, Shoichiro; Oizumi, Akito; Fukushima, Masahiro; Nakayama, Shinsuke

Journal of Nuclear Science and Technology, 21 Pages, 2023/04

 Times Cited Count:6 Percentile:99.12(Nuclear Science & Technology)

The new version of the Japanese evaluated nuclear data library, JENDL-5, was released in December 2021. This paper demonstrates the validation of JENDL-5 for fission reactor applications. Benchmark calculations are performed with the continuous-energy Monte Carlo codes MVP and MCNP and the deterministic code system MARBLE. The benchmark calculation results indicate that the performance of JENDL-5 for fission reactor applications is better than that of the former library JENDL-4.0.

Journal Articles

Development of adjusted nuclear data library for fast reactor application

Yokoyama, Kenji

EPJ Web of Conferences, 281, p.00004_1 - 00004_10, 2023/03

In Japan, development of adjusted nuclear data library for fast rector application based on the cross-section adjustment method has been conducted since the early 1990s. The adjusted library is called the unified cross-section set. The first version was developed in 1991 and is called ADJ91. Recently, the integral experimental data were further expanded to improve the design prediction accuracy of the core loaded with minor actinoids and/or degraded Pu. Using the additional integral experimental data, development of ADJ2017 was started in 2017. In 2022, the latest unified cross-section set AJD2017R was developed based on JENDL-4.0 by using 619 integral experimental data. An overview of the latest version with a review of previous ones will be shown. On the other hand, JENDL-5 was released in 2021. In the development of JENDL-5, some of the integral experimental data used in ADJ2017R were explicitly utilized in the nuclear data evaluation. However, this is not reflected in the covariance data. This situation needs to be considered when developing a unified cross-section set based on JENDL-5. Preliminary adjustment calculation based on JENDL-5 is performed using C/E (calculation/experiment) values simply evaluated by a sensitivity analysis. The preliminary results will be also discussed.

Journal Articles

Japanese Evaluated Nuclear Data Library version 5; JENDL-5

Iwamoto, Osamu; Iwamoto, Nobuyuki; Kunieda, Satoshi; Minato, Futoshi; Nakayama, Shinsuke; Abe, Yutaka*; Tsubakihara, Kosuke*; Okumura, Shin*; Ishizuka, Chikako*; Yoshida, Tadashi*; et al.

Journal of Nuclear Science and Technology, 60(1), p.1 - 60, 2023/01

 Times Cited Count:75 Percentile:99.99(Nuclear Science & Technology)

Journal Articles

Weyl-Kondo semimetal behavior in the chiral structure phase of Ce$$_{3}$$Rh$$_{4}$$Sn$$_{13}$$

Iwasa, Kazuaki*; Suyama, Kazuya*; Kawamura, Seiko; Nakajima, Kenji; Raymond, S.*; Steffens, P.*; Yamada, Akira*; Matsuda, Tatsuma*; Aoki, Yuji*; Kawasaki, Ikuto; et al.

Physical Review Materials (Internet), 7(1), p.014201_1 - 014201_11, 2023/01

 Times Cited Count:2 Percentile:66.84(Materials Science, Multidisciplinary)

Journal Articles

Thinning behavior of solid boron carbide immersed in molten stainless steel for core disruptive accident of sodium-cooled fast reactor

Emura, Yuki; Takai, Toshihide; Kikuchi, Shin; Kamiyama, Kenji; Yamano, Hidemasa; Yokoyama, Hiroki*; Sakamoto, Kan*

Journal of Nuclear Science and Technology, 10 Pages, 2023/00

 Times Cited Count:0 Percentile:0.01(Nuclear Science & Technology)

Journal Articles

Report on the 34th Meeting of Working Party on International Nuclear Data Evaluation Co-operation (WPEC)

Iwamoto, Osamu; Iwamoto, Nobuyuki; Kimura, Atsushi; Tada, Kenichi; Sugawara, Takanori; Yokoyama, Kenji

Kaku Deta Nyusu (Internet), (133), p.1 - 6, 2022/10

no abstracts in English

Journal Articles

Development of evaluation method for core deformation reactivity in sodium-cooled fast reactor; Verification of core deformation reactivity evaluation based on first-order perturbation theory

Doda, Norihiro; Kato, Shinya; Iida, Masaki*; Yokoyama, Kenji; Tanaka, Masaaki

Proceedings of 12th Japan-Korea Symposium on Nuclear Thermal Hydraulics and Safety (NTHAS12) (Internet), 8 Pages, 2022/10

In the conventional core design in sodium-cooled fast reactors (SFRs), negative reactivity feedback due to core deformation was neglected because of large uncertainty in analytical evaluation. To optimize core design, it is necessary to develop an analytical evaluation method and eliminate excessive conservativeness. An evaluation method for core deformation reactivity has been developed by coupling analysis of neutronics, thermal-hydraulics, and structural mechanics. For the verification study of the neutronics calculation method, the reactivity was calculated for the deformed core geometry in which the fuel assembly (FA) moved horizontally in the radial direction for each row from the core center. Compared to reference values derived from Monte Carlo calculations, the calculated reactivity due to FA displacement agreed well in the core region and was overestimated in the reflector region. The modification challenges in development of the core deformation model were identified.

Journal Articles

Development of ARKADIA-Design for design optimization support; Application of coupling method using multi-level simulation technique for plant thermal-hydraulics analysis

Doda, Norihiro; Yoshimura, Kazuo; Hamase, Erina; Yokoyama, Kenji; Uwaba, Tomoyuki; Tanaka, Masaaki

Proceedings of Technical Meeting on State-of-the-art Thermal Hydraulics of Fast Reactors (Internet), 3 Pages, 2022/09

ARKADIA-Design is being developed to support the optimization of sodium-cooled fast reactors in the conceptual design stage. Design optimization requires various types of numerical analysis: 1-D plant dynamics analysis for efficient evaluation of various design options and multi-dimensional analysis for a detailed evaluation of local phenomena, including multi-physics. For those analyses, ARKADIA-Design performs whole plant analyses based on the multi-level simulation (MLS) technique in which the analysis codes are coupled to simulate the phenomena in an intended degree of resolution. This paper describes an outline of the coupling analysis methods in the MLS of the ARKADIA-Design and the numerical simulations of the experimental fast breeder reactor EBR-II tests by the coupled analysis.

Journal Articles

Development of ARKADIA for the innovation of advanced nuclear reactor design process (Overview of optimization process development in design optimization support tool, ARKADIA-Design)

Tanaka, Masaaki; Doda, Norihiro; Yokoyama, Kenji; Mori, Takero; Okajima, Satoshi; Hashidate, Ryuta; Yada, Hiroki; Oki, Shigeo; Miyazaki, Masashi; Takaya, Shigeru

Dai-26-Kai Doryoku, Enerugi Gijutsu Shimpojiumu Koen Rombunshu (Internet), 5 Pages, 2022/07

To assist conceptual studies of various reactor systems conducted by private sectors in nuclear power innovation, development of an innovative design system named ARKADIA (Advanced Reactor Knowledge- and AI-aided Design Integration Approach through the whole plant lifecycle) is undergoing to achieve the design of an advanced nuclear reactor as a safe, economic, and sustainable carbon-free energy source. In this paper, focusing on the ARKADIA-Design as a part of it, the progress in the development of optimization processes on the representative problems in the fields of the core design, the plant structure design, and the maintenance schedule planning are introduced.

Journal Articles

Sodium-cooled Fast Reactors

Ohshima, Hiroyuki; Morishita, Masaki*; Aizawa, Kosuke; Ando, Masanori; Ashida, Takashi; Chikazawa, Yoshitaka; Doda, Norihiro; Enuma, Yasuhiro; Ezure, Toshiki; Fukano, Yoshitaka; et al.

Sodium-cooled Fast Reactors; JSME Series in Thermal and Nuclear Power Generation, Vol.3, 631 Pages, 2022/07

This book is a collection of the past experience of design, construction, and operation of two reactors, the latest knowledge and technology for SFR designs, and the future prospects of SFR development in Japan. It is intended to provide the perspective and the relevant knowledge to enable readers to become more familiar with SFR technology.

Journal Articles

Completion of JENDL-5 and prospects for its application to numerical analysis, 4; Integral test of JENDL-5; Benchmark analysis in fast reactor system

Yokoyama, Kenji; Taninaka, Hiroshi

Kaku Deta Nyusu (Internet), (132), p.25 - 33, 2022/06

This article explains the results of integral test of JENDL-5 by benchmark analysis in fast reactor system, which were presented in a special session of the 2022 Spring Annual Meeting of the Atomic Energy Society of Japan (AESJ). The latest version of Japanese evaluated nuclear data library, JENDL-5, was released in December 2021. In order to confirm the applicability of JENDL-5 to the fast reactor system, we conducted a set of benchmark analysis using the integral experiment data included in the fast reactor nuclear design database which is being developed by JAEA. With respect to major nuclear characteristics in the standard fast reactor system, it was confirmed that the ratios of analysis result and experimental result (C/E values) based on JENDL-5 were almost the same as those of JENDL-4.0. In the special session, the results of sensitivity analysis were reported. Since the results have been described in the proceedings of the AESJ meeting, we add the results of the versions under development of JENDL-5 and discuss their relationship with the reported results of sensitivity analysis.

Journal Articles

Applicability study of Bayesian optimization in core neutronic design using a toy model

Kuwagaki, Kazuki; Yokoyama, Kenji

Proceedings of International Conference on Physics of Reactors 2022 (PHYSOR 2022) (Internet), 10 Pages, 2022/05

In Japan Atomic Energy Agency (JAEA), an innovative design approach named ARKADIA (Advanced Reactor Knowledge- and AI-aided Design Integration Approach through the whole plant lifecycle) for the advanced nuclear reactors is currently under development. One of the tasks in ARKADIA is to build a system that automatically optimizes core and fuel designs by conducting core neutronic and thermal-hydraulic calculations, fuel integrity evaluations, and plant dynamic analyses. This system will be implemented to automatically find an optimal design that minimizes (or maximizes) objective function defined by a core performance, while varying the core and fuel design parameters such as fuel pin diameter, core height and diameter. In this study, as the first step of the system development, we focused only on core neutronic design and conducted a study of automatic optimization. As the optimization algorithm, Bayesian optimization (BO), which is an effective method for optimization problems with expensive computation cost of objective function, was selected. The applicability of BO was studied based on single- and two-objective optimization examples of core neutronic design in a toy model. As a result, in the former, it was confirmed that BO can obtain the optimal solution, which well matches the reference solution calculated by a brute force calculation, with a small number of required calculation executions. Its usability on core neutronic designs, where the computation cost per case is large, was confirmed. In the latter, it was shown that BO can obtain a pareto solutions-set that shows good agreement with the reference solution.

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