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Kawai, Keito*; Nemoto, Yoshiyuki; Fujimura, Yuki; Kondo, Keietsu; Abe, Yosuke; Mohamad, A. B.; Pham, V. H.; Ishikawa, Norito; Ishijima, Yasuhiro; Ioka, Ikuo; et al.
Nihon Genshiryoku Gakkai Wabun Rombunshi (Internet), 24(3), p.82 - 98, 2025/08
An accident tolerant fuel (ATF) cladding, which is more resistant to accidents than the conventional Zircaloy cladding, is under development. One of these claddings is a chromium (Cr)-coated cladding with an outer surface coated with Cr, which is expected to improve the resistance to high-temperature steam oxidation. In this study, electrochemical plating was applied to coat a Cr layer on the cladding outer surface, and its properties under accidental conditions were evaluated. In a loss-of-coolant accident (LOCA), the cladding will burst and both the outer and inner surfaces of the cladding will be oxidized. Thus, as-received Zircaloy-4 and Cr-coated claddings were tested for oxidation in high-temperature steam to investigate differences in oxidation behavior, hydrogen absorption behavior, and mechanical properties after oxidation. Oxidation tests were conducted using a thermobalance. The amount of oxidation of coated samples decreased by half compared with that of uncoated samples, indicating that the coating was effective in inhibiting oxidation. However, the hydrogen absorption of coated samples was found to be higher than that of uncoated samples. In this paper, we discuss the mechanism behind this difference in hydrogen absorption and its effect on mechanical properties.
Otsuka, Satoshi; Tanno, Takashi; Yano, Yasuhide; Kaito, Takeji
Materials and Processes for Nuclear Energy Today and in the Future, p.279 - 297, 2024/10
The oxide dispersion strengthening is an effective technique for improving the mechanical strength of the steel. The dispersed oxides prevent the gliding motion of dislocations, thus remarkably enhancing the resistance to high-temperature deformation and rupture of steels. Extensive efforts have been made to develop ODS steels in the fields of nuclear and fusion engineering. Research has been done to improve their performance and meet the requirements such as irradiation resistance, high-temperature strength, and corrosion resistance. Based on recent research, the high-density dispersion of nanosized oxides could improve the irradiation resistance of the steels in addition to high-temperature strength because the interface between oxide and matrix could act as sink sites for point defects. This section overviews the ODS steel development for nuclear application.
Ishibashi, Ryo*; Hirosaka, Kazuma*; Yamana, Takashi*; Shibata, Masatoshi*; Sasaki, Masana*; Nemoto, Yoshiyuki; Hinoki, Tatsuya*
Proceedings of TopFuel 2024 (Internet), 9 Pages, 2024/09
Nemoto, Yoshiyuki
Kiho Enerugi Sogo Kogaku, 47(1), p.27 - 32, 2024/04
no abstracts in English
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：49.58(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：23 Percentile：77.84(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.
Nemoto, Yoshiyuki; Okada, Yuji*; Sato, Daiki*; Mohamad, A. B.; Ioka, Ikuo; Suzuki, Eriko
no journal, ,
no abstracts in English
Tezuka, Kenichi*; Kino, Chiaki*; Yamashita, Susumu; Mohamad, A. B.; Nemoto, Yoshiyuki
no journal, ,
no abstracts in English
Sugiyama, Tomoyuki
no journal, ,
no abstracts in English
Yamashita, Shinichiro
no journal, ,
A purpose of this presentation is to share the common understanding on the necessity of developing accident tolerant fuel in Japan for enhancing the nuclear safety. This presentation will mainly introduce the R&D results and knowledges obtained under the METI-ATF project since JFY2015.
Yamashita, Shinichiro; Ioka, Ikuo; Nemoto, Yoshiyuki; Kawanishi, Tomohiro; Osaka, Masahiko; Kaji, Yoshiyuki
no journal, ,
After the nuclear accident at Fukushima Daiichi power plant, global interest has expanded in exploring fuels with enhanced performance during severe accident, and enhancing the accident tolerance of light water reactors (LWRs) became a topic of serious discussion all over the world. In Japan, research and development (R&D) program for establishing technical basis of ATF has been conducted by JAEA in cooperation with power plant providers, fuel venders and universities. In this presentation, the overview of ATF R&D program in Japan will be introduced with the explanation on JAEA's role in ATF R&D program.
Yamashita, Shinichiro
no journal, ,
IAEA has organized a Technical Meeting to facilitate the exchange of information regarding evolutionary accident tolerant fuels (eATFs) to enable experts to collaboratively identify the opportunities and challenges faced in all stages of the back end of the fuel cycle (storage, transportation, reprocessing & recycling, and disposal) by the introduction of eATFs into commercial use. Based on the request from IAEA, JAEA will introduce the current status of ATF R&D program in Japan. In addition to that, JAEA will give a perspective of back-end consideration on eATFs.
Kawai, Keito*; Fujimura, Yuki; Kondo, Keietsu; Abe, Yosuke; Ishikawa, Norito; Nemoto, Yoshiyuki; Yamashita, Shinichiro; Ishijima, Yasuhiro; Ioka, Ikuo; Funamoto, Kodai*; et al.
no journal, ,
In the previous report, we reported on the differences in oxidation and hydrogen absorption behavior of Zircaloy-4 cladding specimen and specimen with chromium (Cr) coating on its outer surfaces when they were subjected to oxidation tests in high temperature steam at 1100C for 20 minutes. In this report, we describe the details of the cross-sectional observation of these specimens and the results of the evaluation of the differences in the shape and distribution of hydrides and other substances depending on the presence or absence of Cr coating.
Mohamad, A. B.; Chen, J.*; Ioka, Ikuo; Yamashita, Shinichiro
no journal, ,
no abstracts in English
Ishibashi, Ryo*; Hirosaka, Kazuma*; Yamana, Teppei*; Shibata, Masatoshi*; Sasaki, Masana*; Yasuda, Kenichi*; Nemoto, Yoshiyuki; Hinoki, Tatsuya*
no journal, ,
no abstracts in English
Kawai, Keito*; Nemoto, Yoshiyuki; Fujimura, Yuki; Kondo, Keietsu; Abe, Yosuke; Mohamad, A. B.; Pham, V. H.; Ishikawa, Norito; Ishijima, Yasuhiro; Ioka, Ikuo; et al.
no journal, ,
no abstracts in English
