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Journal Articles

Modeling of the P2M past fuel melting experiments with the FEMAXI-8 code

Mohamad, A. B.; Udagawa, Yutaka

Nuclear Technology, 210(2), p.245 - 260, 2024/02

 Times Cited Count:0 Percentile:0.01(Nuclear Science & Technology)

Journal Articles

Microstructural evolution of intermetallic phase precipitates in Cr-coated zirconium alloy cladding in high-temperature steam oxidation up to 1400$$^{circ}$$C

Mohamad, A. B.; Nemoto, Yoshiyuki; Furumoto, Kenichiro*; Okada, Yuji*; Sato, Daiki*

Corrosion Science, 224, p.111540_1 - 111540_15, 2023/11

Journal Articles

Chemical interaction between Sr vapor species and nuclear reactor core structure

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:0 Percentile:0.01(Nuclear Science & Technology)

Oral presentation

Fuel performance analysis by using FEMAXI code for the fuel samples (HBC4, XM3) burned up to different level

Mohamad, A. B.; Udagawa, Yutaka; Nemoto, Yoshiyuki; Yamashita, Shinichiro

no journal, , 

Oral presentation

Study on Cr coated cladding, 2; Investigation of the metallography after LOCA test

Mohamad, A. B.; Okada, Yuji*; Sato, Daiki*; Ioka, Ikuo; Suzuki, Eriko; Nemoto, Yoshiyuki

no journal, , 

Oral presentation

Study on Cr coated cladding, 1; Investigation of oxidation behavior

Nemoto, Yoshiyuki; Okada, Yuji*; Sato, Daiki*; Mohamad, A. B.; Ioka, Ikuo; Suzuki, Eriko

no journal, , 

no abstracts in English

Oral presentation

Development of analysis method based on SAMPSON for ATF with Cr coated Zr alloy cladding

Tezuka, Kennichi*; Kino, Chiaki*; Yamashita, Susumu; Mohamad, A. B.; Nemoto, Yoshiyuki

no journal, , 

no abstracts in English

Oral presentation

The Behavior of Cr-coated Zry cladding under high-temperature steam oxidation

Mohamad, A. B.; Furumoto, Kenichiro; Nemoto, Yoshiyuki; Ioka, Ikuo; Sato, Daiki*; Okada, Yuji*; Yamashita, Shinichiro; Osaka, Masahiko

no journal, , 

Oral presentation

Overview of ATF R&D program in Japan

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.

Oral presentation

Accident-Tolerant Fuel R&D Program in Japan

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.

Oral presentation

Transition of the Zr(Cr, Fe)$$_{2}$$ intermetallic phase up to the eutectic temperature

Mohamad, A. B.; Nemoto, Yoshiyuki; Furumoto, Kenichiro*; Okada, Yuji*; Sato, Daiki*

no journal, , 

Oral presentation

The Transition of protective coating to no-longer protective coating of Cr-coated Zry cladding in high temperature steam oxidation

Mohamad, A. B.; Nemoto, Yoshiyuki; Furumoto, Kenichiro*; Okada, Yuji*; Sato, Daiki*

no journal, , 

Oral presentation

ATF fundamental research at JAEA

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

Oral presentation

Improving chemical thermodynamics knowledge of severe accidents within the OECD-TCOFF2 Project

Journeau, C.*; Bechta, S.*; Komlev, A.*; Kurata, Masaki; Ohgi, Hiroshi; Matsumoto, Toshinori; Mohamad, A. B.; Barrachin, M.*; Quaini, A.*; Bottomley, D.*; et al.

no journal, , 

Oral presentation

Oxidation behavior of Cr-coated Zry cladding in steam environments

Mohamad, A. B.; Nemoto, Yoshiyuki; Furumoto, Kenichiro*; Okada, Yuji*; Sato, Daiki*

no journal, , 

Oral presentation

Fundamental research program on zircalloy with accident tolerance

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.

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