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

https://doi.org/10.1080/00295450.2023.2185061

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

https://doi.org/10.1080/00223131.2022.2092227

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, , 

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.

Study on coating technic to enhance accident tolerance of fuel cladding, 3; Irradiation behavior of the Cr coated MDA cladding

Mohamad, A. B.; Chien, J.; Ioka, Ikuo; Suzuki, Eriko; Kondo, Keietsu; Nemoto, Yoshiyuki; Okubo, Nariaki; Yamashita, Shinichiro; Okada, Yuji*; Sato, Daiki*

no journal, , 

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.

High-throughput experiments on Cr-rich particle formation in Fe-Cr-Al alloys using combinatorial samples with continuous concentration gradients

Abe, Yosuke; Mohamad, A. B.; Sasaki, Taisuke*; Yamashita, Shinichiro; Okubo, Nariaki; Ukai, Shigeharu

no journal, , 

Embrittlement due to the formation of Cr-rich embrittlement phases ( particles) is a concern in Fe-Cr-Al (ODS) alloys being developed as accident tolerant fuel cladding for light water reactors, but the compositional range of Cr and Al in which embrittlement occurs has not been determined. In this study, combinatorial samples with continuous concentration gradients were prepared by interdiffusion heat treatment of Fe-Cr-Al alloys, and high-throughput composition and hardness measurements were performed before and after thermal aging. The compositional range obtained was sufficient to determine the precipitation boundaries of the particles. Solid solution strengthening was studied from the hardness data after diffusion heat treatment, suggesting that slow cooling leads to spinodal decomposition and formation of Al precipitates. The precipitation boundary of particles was refined from the hardness data after thermal aging.

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.

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, , 

Updating fission product chemistry database based on recent investigation in Fukushima-Daiichi Nuclear Power Station, 1; Overview of fundamental study related to fission product chemistry

Miwa, Shuhei; Nakajima, Kunihisa; Karasawa, Hidetoshi; Rizaal, M.; Luu, V. N.; Mohamad, A. B.

no journal, , 

no abstracts in English