Practical development of accident tolerant FeCrAl-ODS fuel claddings for BWRs in Japan

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.