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Yamashita, Shinichiro
Nihon Genshiryoku Gakkai-Shi ATOMO
, 65(4), p.233 - 237, 2023/04
In the wake of the accident at the Fukushima Daiichi Nuclear Power Plant (NPP) of TEPCO due to the Great East Japan Earthquake in 2011, interest in the early implementation of accident tolerant fuel (ATF) not only for many existing NPPs but also for future NPPs, which is expected to dramatically improve the safety of light water reactors, has increased globally, and research and development is currently underway in many countries around the world. In this article, an overview of domestic ATF technology development that has been carried out with the support of the Ministry of Economy, Trade and Industry since 2015, will be introduced.
Otsuka, Satoshi; Shizukawa, Yuta; Tanno, Takashi; Imagawa, Yuya; Hashidate, Ryuta; Yano, Yasuhide; Onizawa, Takashi; Kaito, Takeji; Onuma, Masato*; Mitsuhara, Masatoshi*; et al.
Journal of Nuclear Science and Technology, 60(3), p.288 - 298, 2023/03
Times Cited Count：4 Percentile：78.52(Nuclear Science & Technology)JAEA has been developing 9Cr-oxide dispersion strengthened (ODS) tempered martensitic steel(TMS) as a candidate material for the fuel cladding tubes of sodium-cooled fast reactors(SFRs). The reliable prediction of in-reactor creep-rupture strength is critical for implementing the 9Cr-ODS TMS cladding tube in the SFR. This study investigated the quantitative correlation between the creep properties of 9Cr-ODS TMS at 700 
C and the dispersions of nanosized oxides by analyzing the creep data and the material's nanostructure. The possibility of deriving a formula for estimating the in-reactor creep properties of 9Cr-ODS TMSs based on an analysis of the nanostructure of neutron-irradiated 9Cr-ODS TMSs was also discussed. The creep properties of 9Cr-ODS TMS at 700 C closely correlated with the dispersion of nanosized oxide particles. The correlation between creep-rupture lives and nanosized oxide particle dispersion was determined using existing creep models. The elucidation of correlation between the stress exponent of secondary creep rate and the nanostructure is essential to enhance future modeling reliability and formulation.
Ukai, Shigeharu*; Yano, Yasuhide; Inoue, Toshihiko; Sowa, Takashi*
Materials Science & Engineering A, 812, p.141076_1 - 141076_11, 2021/04
Times Cited Count：12 Percentile：73.14(Nanoscience & Nanotechnology)FeCrAl oxide dispersion strengthened alloys are promising materials for accident tolerant fuels for light water reactors (LWRs). In these alloys, Al and Cr are key elements with important synergistic effects: enhancement of the formation of oxidation-resistant Al
O
phase by Cr addition and suppression of the formation of the embrittling Cr-rich 
' phase by Al addition. The solid-solution strengthening resulting from Al and Cr co-addition was investigated in this study. The solid-solution strengthening resulting from Al and Cr co-addition was investigated in this study. The Al and Cr contents were systematically varied from 9-16 at.% and 10-17 at.%, respectively, and tensile tests were conducted at 298 K, 573 K and 973 K in the as-annealed condition. The solid solution strengthening increased linearly, 20 MPa per 1 at.% Al and 5 MPa per 1 at.% Cr, at the typical LWR operational temperature of 573 K. The conventional Fleischer-Friedel and Labusch theories cannot explain this level of solid-solution strengthening. It was shown that Suzuki's double kink theory for screw dislocations reasonably predicts the solid solution strengthening by Al and Cr as well as the inverse dependency on the absolute temperature and linear dependency on the Al and Cr content.
Ukai, Shigeharu*; Ono, Naoko*; Otsuka, Satoshi
Comprehensive Nuclear Materials, 2nd Edition, Vol.3, p.255 - 292, 2020/08
Fe-Cr-based oxide dispersion strengthened (ODS) steels have a strong potential for high burnup (long-life) and high-temperature applications typical for SFR fuel cladding. Current progress in the development of Fe-Cr-based ODS steel claddings is reviewed, including their relevant mechanical properties, e.g. tensile and creep rupture strengths in the hoop directions. In addition, this paper reviewed the current research status on corrosion resistant Fe-Cr-Al-based ODS steel claddings, which are greatly paid attention recently as the accident tolerant fuel claddings for the light water reactor (LWR) and also as the claddings of the lead fast reactors (LFR) utilizing Pb-Bi eutectic (LBE) coolant.
Shirasu, Noriko; Saito, Hiroaki; Yamashita, Shinichiro; Nagase, Fumihisa
Proceedings of 2017 Water Reactor Fuel Performance Meeting (WRFPM 2017) (USB Flash Drive), 8 Pages, 2017/09
Silicon carbide (SiC) is an attractive candidate of accident tolerant fuel (ATF) cladding material because of its high chemical stability, high radiation resistance and low neutron absorption. FEMAXI-ATF has been developed to analysis SiC cladding fuel behaviors. The thermal, mechanical and irradiation property models were implemented to FEMAXI-7, which is a fuel behavior analysis code being developed in JAEA. Fuel rod behavior analysis was performed under typical boiling water reactor (BWR) operating conditions with a model based on a 9
9 BWR fuel (Step III Type B), in which the cladding material was replaced from Zircaloy to SiC. The SiC cladding shows large swelling by irradiation. It increases the gap size and decreases cladding thermal conductivity. The mechanism of relaxation of stress is also different from the Zircaloy cladding. The experimental data for SiC materials are still insufficient to construct the models, especially for evaluating fracture behavior.
Yamashita, Shinichiro; Nagase, Fumihisa; Kurata, Masaki; Kaji, Yoshiyuki
Proceedings of Annual Topical Meeting on LWR Fuels with Enhanced Safety and Performance (TopFuel 2016) (USB Flash Drive), p.21 - 30, 2016/09
Fuel rod, channel box, and control rod designed with new materials and concepts have been developed in Japan for increasing accident tolerance of LWRs. In order to efficiently and properly implement the accident tolerant fuels (ATFs) and the other components, it is necessary not only to accumulate fundamental and practical data but also to consider technology readiness, recognize knowledge gaps, and establish strategy for design and fabrication. The Japan Atomic Energy Agency (JAEA) has established the above "technical basis" and drafted a research plan towards implementation of the ATFs and components as a program sponsored and organized by the Ministry of Economy, Trade and Industry (METI). It is useful to take advantage of the experiences in commercial uses of zirconium-base alloys in LWRs and, therefore, JAEA has conducted this METI project in cooperation with power plant providers, fuel venders, research institutes and universities who have been involved in the development of the ATF materials. The present paper describes the main results of the project conducted to establish the technical basis of the ATFs and components.
and SiC in LWR conditionsShirasu, Noriko; Kurata, Masaki
no journal, ,
After Fukushima Daiichi nuclear accident in 2011, enhancing the accident tolerance of light water reactor fuels became a very important issue and currently the development of accident-tolerant fuel (ATF) are in progress. SiC is an attractive candidate of ATF cladding material because of its high chemical stability, high radiation resistance and low neutron absorption. SiC reacts much less than Zircaloy with steam, the generation of heat and hydrogen gas would be extremely suppressed. Thermodynamic evaluation on chemical interactions between UO and SiC were performed on various possible conditions of oxygen potential and temperature in severe accident. SiC is converted to SiO by reaction with O. SiC is also converted to SiO by reaction with HO. However, the fraction of the sub-reaction for forming SiO increases than in the case of reaction with O when comparing the results at the same temperatures and the same oxygen potentials.
Yamashita, Shinichiro; Kondo, Keietsu; Aoki, So; Hashimoto, Naoyuki*; Ukai, Shigeharu*; Sakamoto, Kan*; Hirai, Mutsumi*; Kimura, Akihiko*; Kusagaya, Kazuyuki*
no journal, ,
As the lesson learned from the accident at the Fukushima Daiichi Nuclear Power Station, it is commonly recognized that development of the advanced fuel and core components with enhanced accident tolerance and high reliability is quite important for increasing safety of the existing Light Water Reactors (LWRs). FeCrAl-ODS steel is one of prospective candidate materials with enhanced accident tolerance and needs to be accumulated properly and efficiently fundamental and practical data for core and plant design of nuclear reactor. In this study, hardness measurement and microstructural observation for ion-irradiated FeCrAl-ODS steel were conducted in order to evaluate irradiation property in advance toward a research reactor irradiation test. The results indicated that steep irradiation hardening occurred at the initial stage of irradiation and also that nucleation and growth of irradiation defect cluster occurred at the higher dose than the irradiation hardening occurred.
Yamashita, Shinichiro
no journal, ,
In this oral presentation, the result of "Development of Technical Basis for Introducing Advanced Fuels Contributing to Safety Improvement of Current Light Water Reactors" carried out under the Project on Development of Technical Basis for Safety Improvement at Nuclear Power Plants by Ministry of Economy, Trade and Industry (METI) of Japan will be partially introduced. The gap between the present R&D status and the achievement of practical use in order to progress the R&D properly and efficiently will be talked together with related information of oversea research activities.
Nagase, Fumihisa
no journal, ,
JAEA conducts R&D to support the decommissioning at the Fukushima Daiichi NPS and to contribute improvement of the LWR safety in the frame of domestic and international collaborations as well as the own projects. The R&D mostly focuses on the phenomena in BWRs and covers various issues related to materials degradation in severe accidents. In parallel, JAEA has the research activity to establish technical basis for practical use of accident tolerant fuel (ATF) components in existing LWRs. The preliminary computer code analyses showed necessary material data and subjects to design the ATF components.
Sakamoto, Kan*; Hirai, Mutsumi*; Ukai, Shigeharu*; Kimura, Akihiko*; Yamaji, Akifumi*; Kusagaya, Kazuyuki*; Kondo, Takao*; Ioka, Ikuo; Yamashita, Shinichiro; Kaji, Yoshiyuki
no journal, ,
no abstracts in English
Kakiuchi, Kazuo*; Sato, Hisaki*; Ishibashi, Ryo*; Kondo, Takao*; Ioka, Ikuo; Yamashita, Shinichiro; Kaji, Yoshiyuki
no journal, ,
no abstracts in English
Teshima, Hideyuki*; Watanabe, Seiichi*; Furumoto, Kenichiro*; Kirimura, Kazuki*; Yamakoshi, Yoshinori*; Ioka, Ikuo; Yamashita, Shinichiro; Kaji, Yoshiyuki
no journal, ,
no abstracts in English
Ikegawa, Tomohiko*; Ishibashi, Ryo*; Sakamoto, Kan*; Hirai, Mutsumi*; Yamashita, Shinichiro
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no abstracts in English
Yamashita, Shinichiro; Ioka, Ikuo; Nemoto, Yoshiyuki; Shirasu, Noriko; Kurata, Masaki; Kaji, Yoshiyuki; Fukahori, Tokio; Watanabe, Seiichi*; Kirimura, Kazuki*; Kakiuchi, Kazuo*; et al.
no journal, ,
Since the accident at Fukushima Daiichi Nuclear Power Station, enhancing the accident tolerance of Light Water Reactors (LWRs) became a topic of serious discussion, and the research and development (R&D) for improving the safety LWRs has been activated in many countries. In Japan, the R&D project on accident tolerant fuel and other components (ATFs) of LWRs, which is sponsored and organized by the Ministry of Economy, Trade and Industry (METI), has been initiated in 2015 for establishing technical basis of ATFs. The Japan Atomic Energy Agency (JAEA) has coordinated and carried out this ATF R&D project in cooperation with power plant providers, fuel venders and universities for making the best use of the experiences, knowledges in commercial uses of zirconium-base alloys in LWRs.
Sowa, Takashi*; Ukai, Shigeharu*; Aghamiri, M.*; Shibata, Hironori*; Hayashi, Shigenari*; Ono, Naoko*; Sakamoto, Kan*; Hirai, Mutsumi*; Yamashita, Shinichiro
no journal, ,
no abstracts in English
Hashimoto, Naoyuki*; Toyoda, Kodai*; Sakamoto, Kan*; Hirai, Mutsumi*; Yamashita, Shinichiro
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no abstracts in English
Kimura, Akihiko*; Yuzawa, Sho*; Yabuuchi, Kiyohiro*; Sakamoto, Kan*; Hirai, Mutsumi*; Ukai, Shigeharu*; Yamashita, Shinichiro; Kusagaya, Kazuyuki*
no journal, ,
no abstracts in English
Takahashi, Katsuhito*; Sakamoto, Kan*; Otsuka, Teppei*; Ukai, Shigeharu*; Hirai, Mutsumi*; Yamashita, Shinichiro
no journal, ,
no abstracts in English
Ishibashi, Ryo*; Kondo, Takao*; Yamashita, Shinichiro
no journal, ,
no abstracts in English
