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Aoki, Takeshi; Shimizu, Atsushi; Ishii, Katsunori; Morita, Keisuke; Mizuta, Naoki; Kurahayashi, Kaoru; Yasuda, Takanori; Noguchi, Hiroki; Nomoto, Yasunobu; Iigaki, Kazuhiko; et al.
Annals of Nuclear Energy, 220, p.111503_1 - 111503_7, 2025/09
Times Cited Count：0Aiming to establish coupling technologies between a high temperature gas cooled reactor and a hydrogen production plant, JAEA has initiated the HTTR Heat Application Test Project and is conducting the safety design and the safety analysis for the licensing of the HTTR Heat Application Test Facility. The present study proposed a relative evaluation methodology for the demarcation of applicable laws and design standards for the nuclear hydrogen production system and applied it to the HTTR Heat Application Test Facility. The evaluation results showed that a candidate applying the High Pressure Gas Safety Act to the Heat Application Test Facility (hydrogen production plant) and design standards established under the High Pressure Gas Safety Act to the steam reformer did not show the lowest category in any of the metrics, and was proposed as the most superior demarcation option for the HTTR Heat Application Test Facility.
Wakai, Takashi; Ando, Masanori; Okajima, Satoshi; Toyota, Kodai; Onuma, Terumitsu*; Takahashi, Ryoya*; Asayama, Tai
Dai-29-Kai Doryoku, Enerugi Gijutsu Shimpojiumu Yokoshu (Internet), 5 Pages, 2025/06
This paper describes an experimental study for establishing a passive creep-fatigue test technique that mainly utilizes the difference in thermal expansion coefficients of the materials as material surveillance test technique that can be applied to evaluate the structural integrity of the fast reactor components when the components are used beyond the period assumed in the design. Using the test article designed with the aid of a finite element analysis, a long-term creep-fatigue test data has been successfully obtained. In the designing of the test article, it was essential to generate a adequate strain at the gauge portion of the specimen due to the difference of thermal expansion coefficients of the materials, without buckling. After much trial and error, an optimal shape and dimensions of the test article and the cyclic thermal load conditions are established. In the future, miniaturization of the test article for applying the established test technique to the actual nuclear reactors will be required.
Nagata, Shohei*; Ichida, Toshiyuki*; Fujieda, Daigo; Aoyagi, Kazuhei
Tunnelling into a Sustainable Future; Methods and Technologies; Proceedings of the ITA-AITES World Tunnel Congress 2025 (WTC 2025), p.3517 - 3524, 2025/05
We performed three-dimensional excavation analysis of three shafts and intersection of horizontal tunnel excavated to 500 m depth at Horonobe URL, and evaluated stress acting on concrete lining under various conditions such as differences in diameter of shafts, lining shape, lining thickness, concrete strength, and installation interval of the lining. The design of the support was determined based on the analysis results, and then the construction of the shafts and horizontal galleries were started. Considering the measured stress acting on the concrete lining of the shaft and sprayed mortar installed at the intersection of shaft and horizontal gallery, we could successfully excavate the shaft to 500 m depth and intersection of horizontal gallery at 420 m depth. We also concluded that appropriate support pattern can be selected based on the measured stress acting on the concrete lining and numerical analysis performed prior to the excavation. This can contribute to enhance the reliability of the appropriate observational construction technology targeting excavation of shaft.
Tanaka, Masaaki; Doda, Norihiro; Hamase, Erina; Kuwagaki, Kazuki; Mori, Takero; Okajima, Satoshi; Kikuchi, Norihiro; Yoshimura, Kazuo; Matsushita, Kentaro; Hashidate, Ryuta; et al.
Nihon Kikai Gakkai Rombunshu (Internet), 91(943), p.24-00229_1 - 24-00229_12, 2025/03
To assist conceptual studies of various reactor systems conducted by private sectors in nuclear power innovation, an innovative design system named ARKADIA (Advanced Reactor Knowledge- and AI-aided Design Integration Approach through the whole plant lifecycle) has been developed. In this paper, focusing on the ARKADIA-Design, achievements in the development of optimization processes in the fields of the core design, the plant structure design, and the maintenance schedule planning, as major function of ARKADIA-Design, and numerical analysis methods including coupled analysis to be used for the detailed analysis to confirm the plant performance after optimization are introduced at this point in time.
Maruyama, Shuhei
Robutsuri No Kenkyu (Internet), (78), 7 Pages, 2025/03
no abstracts in English
Hayano, Akira
Genshiryoku Bakkuendo Kenkyu (CD-ROM), 31(2), p.134 - 139, 2024/12
In the design and construction of a repository for high-level radioactive waste, it is considered to set criteria for the rational locating of disposal tunnels and holes, taking into account the effects of faults and fractures distributed in the host rock at the tunnel to pit scale and excavation damaged zones (EDZs) that occur after tunnel excavation on the workability of the disposal tunnels and the long-term stability of the engineered barriers. In addition, tunnel sealing technologies such as tunnel backfilling and hydraulic plugs will be considered to prevent short-circuit flow of radionuclides through the excavated tunnels and the surrounding EDZs. In Task B of the HIP, we will attempt to establish criteria for the layout of disposal tunnels and holes, and the systematic integration of technical options related to the design, construction, operation and closure of the repository, using the investigation and construction at the Horonobe URL for Neogene sedimentary rocks as a case study. Based on information provided by the participating organisations and existing case studies from Japan and overseas, the current focus is on the reduction of the mechanical strength of the rock mass, the effects of water inflow on disposal tunnels and holes, or the effects on engineered barriers, in relation to the locating of disposal tunnels and boreholes and the emplacement of engineered barriers, and is currently working on analyses to predict features associated with these. Investigation, construction and testing of the 500 m niches No. 8 and No. 9 to confirm the validity of these predictions, and full-scale tunnel backfilling and hydraulic plug installation test in the 350 m Niche No. 6, will be carried out in Phase 2 of the HIP.
Aoki, Takeshi; Hasegawa, Takeshi; Kurahayashi, Kaoru; Nomoto, Yasunobu; Shimizu, Atsushi; Sato, Hiroyuki; Sakaba, Nariaki
Proceedings of 11th International Topical Meeting on High Temperature Reactor Technology (HTR 2024), 6 Pages, 2024/10
Japan Atomic Energy Agency (JAEA) is planning to perform a test named HTTR heat application test coupling HTTR (High temperature engineering test reactor) and a hydrogen production plant. The present study reports results of the safety design and safety analysis for HTTR heat application test facility. As a safety design, safety classification of structures, systems, and components was defined in the test facility based on their safety functions. As a preliminary safety analysis, a thermal-hydraulic analysis was performed with RELAP5 code. The safety analysis revealed that newly identified events for HTTR heat application test facility except for the rupture of heat transfer tube of steam generator was enveloped by the licensing basis events in conventional HTTR. The preliminary analysis proved that the safety criteria is satisfied in the candidate of licensing basis event.
Iwamura, Toko; Nakata, Hisakazu; Maekawa, Keisuke; Sakai, Akihiro; Sakamoto, Yoshiaki
JAEA-Review 2024-032, 39 Pages, 2024/08
JAEA-Review-2024-032.pdf:6.5MB
Japan Atomic Energy Agency (JAEA) is responsible for the disposal of low-level radioactive waste generated by JAEA itself and research facilities under the revised JAEA Act of 2008 and subsequently developed a "Plan for the Implementation of Disposal Operations" (implementation plan) in 2009. Furthermore, based on the results of the survey on the amount of waste generated by research facilities, the quantity of wastes for the near surface disposal was set at 600,000 in terms of 200L drums, and the results of the consideration on the conceptual design of the disposal facility were summarized in 2012. In 2018 JAEA published its long-term outlook and policy regarding back-end measures in "Back-end Roadmap", and in this "Back-end Roadmap", the amount of waste generated by JAEA was also organized and published. Therefore, the amount of waste materials from waste generators outside JAEA was re-examined, and as a result, the size of the burial facility was changed from 600,000 to 750,000 in terms of 200L drums, and approval was obtained for a change in the implementation plan. In addition, the conceptual design of the disposal facility was revised to accommodate the increased size of the facility. This report summarizes the results of the updated assumptions and disposal facility design from the 2012 conceptual design.
Tanaka, Masaaki; Doda, Norihiro; Hamase, Erina; Kuwagaki, Kazuki; Mori, Takero; Okajima, Satoshi; Kikuchi, Norihiro; Yoshimura, Kazuo; Matsushita, Kentaro; Hashidate, Ryuta; et al.
Dai-28-Kai Doryoku, Enerugi Gijutsu Shimpojiumu Koen Rombunshu (Internet), 5 Pages, 2024/06
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. In this paper, focusing on the ARKADIA-Design, achievements in the development of optimization processes in the fields of the core design, the plant structure design, and the maintenance schedule planning, as major function of ARKADIA-Design, and numerical analysis methods to be used for the detailed analysis to confirm the plant performance after optimization are introduced at this point in time.
Yamano, Hidemasa; Takano, Kazuya; Kurisaka, Kenichi; Kikuchi, Shin; Kondo, Toshiki; Umeda, Ryota; Sato, Rika; Shirakura, Shota*
Dai-28-Kai Doryoku, Enerugi Gijutsu Shimpojiumu Koen Rombunshu (Internet), 5 Pages, 2024/06
This project studies investigation on safety design guideline and risk assessment technology for sodium-cooled fast reactor with the molten-salt heat storage system, development of evaluation method for heat transferring performance between sodium and molten-salt and improvement of the performance, and evaluation of chemical reaction characteristic between sodium and molten-salt and improvement of its safety. This paper describes the effect of sodium-molten salt heat transfer tube failure in addition to the project overview and progress.
Uchibori, Akihiro; Okano, Yasushi
Isotope News, (793), p.32 - 35, 2024/06
The design of a containment vessel in a sodium-cooled fast reactor was optimized from simulation on the hypothetical severe accident including sodium leakage and combustion. The simulation method is one of the base technologies of the design optimization system, ARKADIA. The simulation was performed on the different design conditions including volume of the containment vessel and the safety equipment as optimization parameters. The iterative simulation successfully found that the safety under this accident was kept even in the downsized containment vessel by selecting an effective safety equipment. This study demonstrated that the developed method has basic capability for design optimization in ARKADIA.
Doda, Norihiro; Nakamine, Yoshiaki*; Yoshimura, Kazuo; Kuwagaki, Kazuki; Hamase, Erina; Yokoyama, Kenji; Kikuchi, Norihiro; Mori, Takero; Hashidate, Ryuta; Tanaka, Masaaki
Keisan Kogaku Koenkai Rombunshu (CD-ROM), 29, 6 Pages, 2024/06
As a part of the development of the "Advanced Reactor Knowledge- and AI-aided Design Integration Approach through the whole plant lifecycle (ARKADIA)" to utilize the knowledge obtained through the sodium-cooled fast reactors (SFRs) and combine the latest numerical simulation technologies, ARKADIA-Design is being developed to support the optimization of SFRs in the conceptual design stage. ARKADIA-Design consists of three systems of Virtual Plant Life System (VLS), Enhanced and AI-aided optimization System (EAS), and Knowledge Management System (KMS). A design optimization framework controls the linkage among the three systems through the interfaces in each system. In this study, we have developed a prototype of the framework for core design optimization using the coupled analysis functions in VLS and optimization control function in the linkage of EAS and VLS to investigate the applicability of the framework to the SFR design optimization process.
Kamide, Hideki; Asayama, Tai; Wakai, Takashi; Ezure, Toshiki; Uchibori, Akihiro; Kubo, Shigenobu; Takeuchi, Masayuki
Nuclear Engineering and Design, 421, p.113062_1 - 113062_10, 2024/05
Times Cited Count：0 Percentile：0.00(Nuclear Science & Technology)A sodium cooled fast reactor (SFR) is one of the most relevant and decarbonized energy supply system with higher sustainability on natural resources, footprint, and waste management. It was planned in a strategic roadmap of fast reactor decided by Inter-Ministerial Council for Nuclear Power Japan in 2022 to start a conceptual design of a demonstration reactor from 2024 with a background of accumulated knowledge and experiences of SFR development. For example, a design and lifecycle simulation/evaluation system named ARKADIA has been developed to accelerate such design works. It will enable to take into account plant lifecycle, e.g., operation and maintenance, to the plant design and optimize it based on simulations and knowledgebase. This paper shows research progresses of ARKADIA, safety design and evaluations, codes and standards, fuel cycle, and SFR development projects in Japan.
Tanaka, Masaaki; Enuma, Yasuhiro; Okano, Yasushi; Uchibori, Akihiro; Yokoyama, Kenji; Seki, Akiyuki; Wakai, Takashi; Asayama, Tai
Mechanical Engineering Journal (Internet), 11(2), p.23-00424_1 - 23-00424_13, 2024/04
The outline and development status of element functions and design optimization process in ARKADIA to transform advanced nuclear reactor design to meet expectations of a safe, economic, and sustainable carbon-free energy source are introduced. It is also briefly explained that ARKADIA will realize Artificial Intelligence (AI)-aided integrated numerical analysis to offer the best possible solutions for the design and operation of a nuclear plant including optimization of safety equipment, and merge state-of-the-art numerical simulation technologies and a knowledge base that stores data and insights from past nuclear reactor development projects and R&Ds with AI technologies.
Nishida, Akemi
Doboku Gakkai Dai-14-Kai Kozobutsu No Shogeki Mondai Ni Kansuru Shinpojiumu Rombunshu (Internet), 5 Pages, 2024/01
no abstracts in English
Yamano, Hidemasa; Kurisaka, Kenichi; Takano, Kazuya; Kikuchi, Shin; Kondo, Toshiki; Umeda, Ryota; Shirakura, Shota*
Dai-27-Kai Doryoku, Enerugi Gijutsu Shimpojiumu Koen Rombunshu (Internet), 5 Pages, 2023/09
This project studies investigation on safety design guideline and risk assessment technology for sodium-cooled fast reactor with the molten-salt heat storage system, development of evaluation method for heat transferring performance between sodium and molten-salt and improvement of the performance, and evaluation of chemical reaction characteristic between sodium and molten-salt and improvement of its safety. The project overview is presented in this report.
Okuda, Takahiro; Takahashi, Hideki*; Watakabe, Tomoyoshi
Mechanical Engineering Journal (Internet), 10(4), p.23-00075_1 - 23-00075_9, 2023/08
In recent years, to make the seismic design more rational for the piping systems in nuclear power plants, it has been expected to develop a design method considering plastic deformation and the accompanying energy dissipation of the piping itself. In this study, an extensive series of seismic response analyses was conducted to investigate the degree of influence of the plastic deformation of the pipe support structures on the seismic response of the entire piping system. The analyses include; plasticity is considered for (1) none, (2) the piping only, (3) the support structure only, and (4) both the piping and the support structure.
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.
Kikuchi, Norihiro; Imai, Yasutomo*; Yoshikawa, Ryuji; Doda, Norihiro; Tanaka, Masaaki
Journal of Nuclear Engineering and Radiation Science, 9(3), p.031401_1 - 031401_11, 2023/07
In the design study of an advanced sodium-cooled fast reactor (Advanced-SFR) investigated in JAEA, the use of a specific fuel assembly with an inner duct structure called FAIDUS has been investigated to enhance safety of Advanced-SFR. Since the fuel rods have an asymmetric layout by the inner duct, the validity confirmation of the numerical results of an in-house subchannel analysis code named ASFRE was required. In this paper, therefore, the code-to-code comparisons was applied with numerical results of ASFRE and those of an in-house CFD code named SPIRAL. The applicability of ASFRE was indicated through the confirmation of the consistency of specific temperature distributions.
Doda, Norihiro; Nakamine, Yoshiaki*; Kuwagaki, Kazuki; Hamase, Erina; Kikuchi, Norihiro; Yoshimura, Kazuo; Matsushita, Kentaro; Tanaka, Masaaki
Keisan Kogaku Koenkai Rombunshu (CD-ROM), 28, 5 Pages, 2023/05
As a part of the development of the "Advanced Reactor Knowledge- and AI-aided Design Integration Approach through the whole plant lifecycle (ARKADIA)" to automatically optimize the life cycle of innovative nuclear reactors including fast reactors, ARKADIA-design is being developed to support the optimization of fast reactor in the conceptual design stage. ARKADIA-Design consists of three systems (Virtual plant Life System (VLS), Evaluation assistance and Application System (EAS), and Knowledge Management System (KMS)). A design optimization framework controls the connection between the three systems through the interfaces in each system. This paper reports on the development of an optimization analysis control function that performs design optimization analysis combining plant behavior analysis by VLS and optimization study by EAS.
