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Mohamad, A. B.; Chen, J.*; Ioka, Ikuo*; Suzuki, Eriko; Kondo, Keietsu; Abe, Yosuke; Yamashita, Shinichiro; Okubo, Nariaki; Nemoto, Yoshiyuki; Okada, Yuji*; et al.
Journal of Nuclear Materials, 625, p.156513_1 - 156513_9, 2026/04
Times Cited Count：0 Percentile：0.00(Materials Science, Multidisciplinary)Mohamad, A. B.; Yamashita, Shinichiro; Nemoto, Yoshiyuki; Abe, Yosuke; Pham, V. H.; Ioka, Ikuo; Soma, Yasutaka; Ishijima, Yasuhiro; Sato, Tomonori; Rizaal, M.; et al.
Proceedings of TopFuel 2025; Nuclear Reactor Fuel Performance Conference (Internet), 8 Pages, 2025/10
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.
Hirooka, Shun; Horii, Yuta; Hayashizaki, Kohei; Mohamad, A. B.
Kaku Nenryo, (60-1), p.17 - 20, 2025/02
no abstracts in English
Mohamad, A. B.; Soma, Yasutaka; Nemoto, Yoshiyuki; Abe, Yosuke; Ioka, Ikuo; Sato, Tomonori; Ishijima, Yasuhiro; Miwa, Shuhei; Nakajima, Kunihisa; Yamashita, Shinichiro; et al.
Proceedings of TopFuel 2024 (Internet), 8 Pages, 2024/10
Mohamad, A. B.; Udagawa, Yutaka
Nuclear Technology, 210(2), p.245 - 260, 2024/02
Times Cited Count：2 Percentile：28.31(Nuclear Science & Technology)
CMohamad, A. B.; Nemoto, Yoshiyuki; Furumoto, Kenichiro*; Okada, Yuji*; Sato, Daiki*
Corrosion Science, 224, p.111540_1 - 111540_15, 2023/11
Times Cited Count：12 Percentile：63.59(Materials Science, Multidisciplinary)Mohamad, A. B.; Nakajima, Kunihisa; Miwa, Shuhei; Osaka, Masahiko
Journal of Nuclear Science and Technology, 60(3), p.215 - 222, 2023/03
Times Cited Count：1 Percentile：9.95(Nuclear Science & Technology)Mohamad, A.*; Nakajima, Kunihisa; Suzuki, Eriko; Miwa, Shuhei; Osaka, Masahiko; Oishi, Yuji*; Muta, Hiroaki*; Kurosaki, Ken*
Proceedings of International Topical Workshop on Fukushima Decommissioning Research (FDR 2019) (Internet), 4 Pages, 2019/05
In the accident of Fukushima Daiichi Nuclear Power Station, formation of a volatile SrCl
could have occurred by the sea-water injection into the core. This can cause the release of non-volatile group Sr from the fuel to induce chemical reactions with reactor structural materials, such as stainless steel and Zircaloy (Zry) cladding. Such reactions could cause the changes in distribution of Sr in the reactor. Chemical reactions between Sr species and Zry were therefore investigated experimentally. As the result, it can be said that Sr vapor species were chemically trapped right after the release from fuel. This trapping effect of Sr by Zry-cladding implies a possibility of preferable Sr retention in the oxide phase of debris.
Mohamad, A. B.; Soma, Yasutaka; Nemoto, Yoshiyuki; Abe, Yosuke; Ioka, Ikuo; Sato, Tomonori; Ishijima, Yasuhiro; Miwa, Shuhei; Nakajima, Kunihisa; Kaji, Yoshiyuki; et al.
no journal, ,
Japan Atomic Energy Agency (JAEA) has launched fundamental researches on zircalloy with accident tolerance since 2019. The main purposes of the fundamental researches are to deepen the understanding of the zircalloy behavior under long-term normal operation or Loss of Coolant Accident (LOCA), beyond design basis accident (B-DBA) and severe accident (SA) conditions, and to support the implementation of Cr-coated zircalloy which is being developed by Japanese vendor. JAEA has also been conducted basic technology developments which is necessary for the understanding of the behavior of accident tolerant coated-zircalloy under normal operation, LOCA, B-DBA and SA conditions. For example, the ion irradiation technique combined with light water reactor (LWR) coolant conditions is being developed to simulate the normal operation condition. In addition, to understand LOCA phenomena, the results obtained from the LOCA test are implemented in the machine learning to understand in more detail the cladding fracture and ballooning. Furthermore, a separate effect test, such as the high temperature oxidation test, is also carried out. The fission product release during the B-DBA and SA are also included in the research program. The research results obtained by using these basic technologies will be integrated and implemented into the fuel performance analysis code to predict the fuel performance under reactor operating conditions.
Yamashita, Shinichiro; Mohamad, A. B.; Soma, Yasutaka; Nemoto, Yoshiyuki; Ioka, Ikuo; Kaji, Yoshiyuki; Osaka, Masahiko
no journal, ,
Japan's Accident Tolerant Fuel (ATF) research and development (R&D) program has been conducted since 2015 in cooperation with power plant providers, fuel venders and universities for making the most use of the experiences in R&D, practical design, and evaluations of fuels and cores of commercial Light Water Reactors (LWRs). An overview of the present R&D progress is given, including the role of Japan Atomic Energy Agency (JAEA) in the program.
Miwa, Shuhei; Nakajima, Kunihisa; Karasawa, Hidetoshi; Rizaal, M.; Luu, V. N.; Mohamad, A. B.
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
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
Mohamad, A. B.; Furumoto, Kenichiro; Nemoto, Yoshiyuki; Ioka, Ikuo; Sato, Daiki*; Okada, Yuji*; Yamashita, Shinichiro; Osaka, Masahiko
no journal, ,
Yamashita, Shinichiro; Mohamad, A. B.; 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 including the recent result from fundamental ATF studies.
Yamashita, Shinichiro; Mohamad, A. B.; Ioka, Ikuo; Nemoto, Yoshiyuki; Kawanishi, Tomohiro; Kaji, Yoshiyuki; Osaka, Masahiko; Murakami, Nozomu*; Owaki, Masao*; Sasaki, Masana*; et al.
no journal, ,
Japan's Accident Tolerant Fuel (ATF) research and development (R&D) program has been conducted since 2015 in cooperation with power plant providers, fuel venders and universities for making the most use of the experiences in R&D, practical design, and evaluations of fuels and cores of commercial Light Water Reactors (LWRs). An overview of the present R&D progress is given, in relation to the role of Japan Atomic Energy Agency (JAEA) in the program. The ATF candidate materials currently under consideration are the following three claddings: the silicon carbide (SiC) composite which is potentially applicable for Pressurized Water Reactor (PWR) and Boiling Water Reactor (BWR), the FeCrAl steel strengthened by dispersion of fine oxide particles (FeCrAl-ODS) for BWR, and Cr-coated zircalloy claddings for PWR. In addition to the cladding materials, R&D on the SiC-made BWR channel box and accident tolerant control rods are also underway.
Mohamad, A. B.; Nemoto, Yoshiyuki; Soma, Yasutaka; Ishijima, Yasuhiro; Sato, Tomonori; Ioka, Ikuo; Pham, V. H.; Miwa, Shuhei; Nakajima, Kunihisa; Kaji, Yoshiyuki; et al.
no journal, ,
no abstracts in English
Journeau, C.*; Bechta, S.*; Komlev, A.*; Kurata, Masaki; Ohgi, Hiroshi; Matsumoto, Toshinori; Mohamad, A. B.; Barrachin, M.*; Quaini, A.*; Bottomley, D.*; et al.
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