Initialising ...
Initialising ...
Initialising ...
Initialising ...
Initialising ...
Initialising ...
Initialising ...
Sakamoto, Kan*; Sakaguchi, Chisato*; Miura, Yusuke*; Yokoyama, Hironori*; Matsunaga, Junji*; Kasahara, Hideyuki*; Miyata, Hajime*; Ioka, Ikuo; Yamashita, Shinichiro; Osaka, Masahiko
Proceedings of 2023 Water Reactor Fuel Performance Meeting (WRFPM 2023), p.20 - 28, 2024/00
An oxide-dispersion-strengthened FeCrAl (FeCrAl-ODS) has been continuously developed in Japan as a promising candidate alloy for the accident tolerant fuel cladding of BWRs (boiling water reactors). This paper will introduce the progress in practical development of accident tolerant FeCrAl-ODS fuel claddings for BWRs in the program fully or partially supported and organized by the Ministry of Economy, Trade and Industry (METI) of Japan. The experimental studies have been conducted to obtain and accumulate key material properties of FeCrAl-ODS fuel claddings to support the evaluations in the analytical studies. For the evaluation at normal operation condition, fatigue test of unirradiated fuel cladding and tensile test of irradiated sheet specimen were conducted. In the fatigue test, a tensile-compressive bending strain was loaded on the C-shaped specimens by cyclic movement of a push-pull rod. Test temperature was 623 K, frequency was 1 Hz, and strain amplitude were 0.27, 0.34 and 0.55 %. The results of fatigue tests demonstrated that cycles to failure of the FeCrAl-ODS cladding were higher than that of the O'Donnell and Langer fatigue curve of Zr-based alloy. The tensile test was conducted in a hot cell using the SS-J2 type specimens at ambient temperature, 573 K and 623 K at a strain rate of 10-3 s-1. The specimens were irradiated up to 7.8 and 13 dpa at 573 K in the High Flux Isotope Reactor at ORNL. The irradiation hardening and ductility loss obtained at 7.8 and 13 dpa were comparable to those at 3.9 dpa.
Osaka, Masahiko
no journal, ,
Overview of R&D on nuclear fuels conducted in Japan are reviewed. Fuel safety research at Japan Atomic Energy Agency (JAEA) as the technical support organization for the nuclear regulation consists of those covering normal operation to the beyond design-basis accident. A fundamental study on fission product (FP) chemistry is also being conducted in JAEA towards the improved source term both for the LWR safety enhancement and 1F decommissioning issues. An innovative nitride fuel is being investigated, considering the wider application as a high-performance fuel. Study in Central Research Institute of Electric Power Industry on the fuel degradation and relocation is mentioned. Various fundamental studies in Japanese universities on fuel, cladding and debris/FPs have been conducted.
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.