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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.
Toyota, Kodai; Onizawa, Takashi; Wakai, Eiichi*
Research & Development in Material Science (Internet), 21(5), p.2632 - 2637, 2025/06
Sakurai, Junya*; Torigata, Keisuke*; Matsunaga, Manabu*; Takanashi, Naoto*; Hibino, Shinya*; Kizu, Kenichi*; Morita, Akira*; Inomoto, Masahiro*; Shimohata, Nobuaki*; Toyota, Kodai; et al.
Tetsu To Hagane, 111(5), p.246 - 262, 2025/04
Imagawa, Yuya; Toyota, Kodai; Onizawa, Takashi; Kato, Shoichi
JAEA-Data/Code 2024-010, 90 Pages, 2024/11
JAEA-Data-Code-2024-010.pdf:5.41MB
To establish a material testing technique in sodium and to develop a method to evaluate the sodium environmental effects, sodium tests on fast reactor fuel cladding have been carried out. Fast reactor fuel cladding is susceptible to corrosion thinning and compositional change due to sodium because of its high temperature (around 675
C) and thin wall (about 0.5 mm) during normal operation. Therefore, it is important to evaluate the corrosion behavior and mechanical properties under a high-temperature sodium environment. This report summarizes the results of experimental studies on corrosion behavior and mechanical properties of modified type 316 stainless steel fuel cladding applied to actual fast reactors under a high-temperature sodium environment, in order to reflect the results to future research activities and to consolidate knowledge and experience.
Imagawa, Yuya; Toyota, Kodai; Onizawa, Takashi; Kato, Shoichi
JAEA-Testing 2023-004, 76 Pages, 2024/03
JAEA-Testing-2023-004.pdf:2.08MB
This manual describes the methods for conducting material tests in air, argon gas, and sodium, and for organizing the data obtained, as a part of the development of high-temperature structural design technology for fast reactors. This manual reflects the revision of test methods in Japanese Industrial Standards (JIS) to the "FBR Metallic Materials Test Manual, PNC TN241 77-03" published in 1977 and the "FBR Metallic Materials Test Manual (Revised Edition), JNC TN9520 2001-001" published in 2001. Also, it was written with reference to the recommended room temperature / elevated temperature tensile test method by the Japan Society of Mechanical Engineers (JSME) and the test standard for the elevated-temperature low-cycle fatigue test method by the Society of Materials Science, Japan (JSMS), which are the standard for material test methods in the domestic academic society.
C in Mod.9Cr-1Mo steel to 1
10
cyclesToyota, Kodai; Imagawa, Yuya; Onizawa, Takashi; Kato, Shoichi; Furuya, Yoshiyuki*
Nihon Kikai Gakkai Rombunshu (Internet), 89(928), p.23-00206_1 - 23-00206_15, 2023/12
In order to design fast reactors, it is necessary to consider high cycle fatigue of structural materials up to 110
cycles; to evaluate high cycle fatigue at 110 cycles, it is necessary to develop a best-fit fatigue curve applicable up to 110 cycles. In this study, high cycle fatigue tests were conducted under strain-controlled conditions and ultrasonic fatigue tests were also conducted to develop a high cycle fatigue evaluation method for Mod.9Cr-1Mo steel, which is a candidate material for fast reactor structural materials. Based on the test results, the best-fit fatigue curves were extended and the applicability of the JSME best-fit fatigue curves up to 110 cycles was verified.
Van Rooyen, I. J.*; Ivan, L.*; Messner, M.*; Edwards, L.*; Abonneau, E.*; Kamiji, Yu; Lowe, S.*; Nilsson, K.-F.*; Okajima, Satoshi; Pouchon, M.*; et al.
Proceedings of 4th International Conference on Generation IV and Small Reactors (G4SR-4), p.2 - 12, 2022/10
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.
Toyota, Kodai; Onizawa, Takashi; Wakai, Takashi; Hashidate, Ryuta; Kato, Shoichi
Proceedings of International Conference on Fast Reactors and Related Fuel Cycles; Sustainable Clean Energy for the Future (FR22) (Internet), 10 Pages, 2022/04
Toyota, Kodai; Hashidate, Ryuta; Takahashi, Keita; Yada, Hiroki; Takaya, Shigeru
Hozengaku, 20(2), p.95 - 103, 2021/07
Ando, Masanori; Toyota, Kodai; Hashidate, Ryuta; Onizawa, Takashi
Proceedings of ASME 2021 Pressure Vessels and Piping Conference (PVP 2021) (Internet), 10 Pages, 2021/07
The ASME Boiler and Pressure Vessel Code (ASME BPVC) Section III, Division 5 had provided only one design fatigue curve for Grade 91 steel (Gr.91) at 540 C until 2019 version. To overcome this disadvantage, The ASME Section III Working Group had taken an action to incorporate the temperature-dependent design fatigue curves for Gr. 91 developed by Japan Society of Mechanical Engineers into ASME BPVC Section III Division 5. As the results, the temperature dependent design fatigue curves are provided in the 2021 edition of the ASME BPVC. To clear the features of the best fit fatigue curve equation, 305 data stored in the database were analyzed and the statistic values and the values of 95% and 99% lower confidence bound calculated by failure probability assessment were clarified. Moreover, some additional available data of fatigue and creep-fatigue test obtained in Japan are also indicated for considering the creep-fatigue damage evaluation under elevated temperature condition.
Hashidate, Ryuta; Toyota, Kodai; Takahashi, Keita; Yada, Hiroki; Takaya, Shigeru
Hozengaku, 19(4), p.115 - 122, 2021/01
In order to improve both safety and economic efficiency of a nuclear power plant, it is necessary to realize rational maintenance based on characteristics of the plant. The prototype fast-breeder reactor in Japan, Monju, spent most of the year for the maintenance. Thus, it is important to identify causes of the prolonged maintenance of Monju and to investigate countermeasures for implementation of rational maintenance of next-generation fast reactors. In this study, the authors investigated the causes of the prolonged maintenance of Monju during reactor cold shutdown based on the plant schedule of Monju. In addition, we proposed the maintenance optimization idea for next-generation fast reactors to solve the revealed issues.
Toyota, Kodai
no journal, ,
no abstracts in English
Onizawa, Takashi; Toyota, Kodai; Imagawa, Yuya; Okajima, Satoshi; Ando, Masanori
no journal, ,
In order to realize a fast reactor that achieves both safety and economic efficiency at a high level, Japan Atomic Energy Agency (JAEA) is developing the material strength standard for fast reactor design. JAEA has developed the material strength standard based on the acquired data and its evaluation results, and the standard have been incorporated in the Japan Society of Mechanical Engineers (JSME) code, Rules on the Design and Construction of Nuclear Power Plants, Section II, Fast Reactors (JSME D&C FRs Code). This paper describes the standard that recently incorporated in the JSME D&C FRs code and ongoing studies for improvements in the near future.
Toyota, Kodai; Hashidate, Ryuta; Yada, Hiroki; Takaya, Shigeru; Miyakoshi, Hiroyuki; Kato, Shoichi
no journal, ,
Probabilistic fracture mechanics (PFM) evaluation requires information on the probabilistic distributions of the number and size of initial defects, material properties such as crack growth rate due to fatigue and creep, and load to evaluate the failure probability of components. In this study, ultrasonic testing was conducted on the welds of the test equipment used in the research and development of fast reactors, and the number and size of defects were evaluated. The results could be used as conservative values of initial defects, and the values related to initial defects for PFM evaluation of FBR components were examined.
Wakai, Eiichi; Iwamoto, Yosuke; Shibayama, Tamaki*; Sato, Koichi*; Toyota, Kodai; Onizawa, Takashi; Wakui, Takashi; Ishida, Taku*; Makimura, Shunsuke*; Nakagawa, Yuki*; et al.
no journal, ,
In the fields of accelerator target systems, nuclear power, aerospace, etc., radiation degradation of structural materials and equipment occurs, and therefore, the development of materials with high durability and excellent functions is expected. In order to create innovative materials that can be used in radiation fields, we are developing a new non-destructive inspection technique that can accurately measure the internal defects of various materials in radiation fields. As an innovative material, high-entropy alloys (HEA) are known for their high strength and ductility, and are expected to be used in various applications. In this talk, we will report on the construction of a measurement principle that enables multi-simultaneous measurements even in radiation fields, the status of HEA prototypes, and the status and progress of irradiation analysis of metals and other materials.
Toyota, Kodai; Imagawa, Yuya; Onizawa, Takashi
no journal, ,
no abstracts in English
Imagawa, Yuya; Toyota, Kodai; Onizawa, Takashi
no journal, ,
no abstracts in English
Hashidate, Ryuta; Toyota, Kodai; Takahashi, Keita; Yada, Hiroki; Takaya, Shigeru
no journal, ,
In order to achieve both safety and economic efficiency of a nuclear power plant, it is necessary to realize rational maintenance based on characteristics of the plant. The fast breeder prototype reactor, Monju, spent most of the year for the maintenance. It is important to identify causes of the prolonged maintenance of Monju and to investigate countermeasures for implementation of rational maintenance of a next fast reactor. In this study, causes of the prolonged maintenance of Monju during reactor cold shutdown were investigated based on the plant operation of Monju. In addition, proposals for the maintenance optimization idea of a next generation fast reactor were presented to address the revealed issues.
Toyota, Kodai; Hashidate, Ryuta; Takahashi, Keita; Yada, Hiroki; Takaya, Shigeru
no journal, ,
In order to achieve both safety and economic efficiency of a nuclear power plant, it is necessary to realize rational maintenance based on characteristics of the plant. The fast breeder prototype reactor, Monju, spent most of the year for the maintenance. It is important to identify causes of the prolonged maintenance of Monju and to investigate countermeasures for implementation of rational maintenance of a next fast reactor. In this study, causes of the prolonged maintenance of Monju during reactor cold shutdown were investigated based on the plant maintenance plan of Monju. In addition, proposals for the maintenance optimization idea of a next generation fast reactor were presented to address the revealed issues.
