Development of accident tolerant FeCrAl-ODS fuel cladding for BWRs in Japan

K. Sakamoto*, Y. Miura, S. Ukai, N. H. Oono, A. Kimura, A. Yamaji, K. Kusagaya, S. Takano, T. Kondo, T. Ikegawa, I. Ioka, S. Yamashita


研究成果: Article査読

11 被引用数 (Scopus)


The FeCrAl-oxide dispersion strengthened (ODS) alloy is a promising candidate alloy for the accident tolerant fuel (ATF) cladding of light water reactors (LWRs) and has been recently developed in Japan. This paper reports on the progress of the development of accident tolerant FeCrAl-ODS fuel claddings for boiling water reactors (BWRs) in Japan. Both experimental and analytical studies were conducted to evaluate the influence of implementation of the FeCrAl-ODS fuel cladding to the current BWRs. In the experimental study, key material properties of FeCrAl-ODS fuel cladding have been obtained and accumulated using bar, sheet and tube-shaped materials to support the evaluations in the analytical study. In the evaluation at normal operating conditions, strength and ductility, corrosion property, tritium permeability, wear property, resistance to iodine stress corrosion cracking (SCC) and weldability were examined. A preliminary assessment of compatibility of the FeCrAl-ODS fuel cladding with the current recycling system in Japan was also conducted. In the evaluation of the design basis accident and the beyond design basis accident, strength, steam oxidation property, resistance to water quenching during a loss-of-coolant accident (LOCA) and the LOCA burst property were examined. In order to evaluate the influence of implementation of the FeCrAl-ODS cladding to the current BWRs, the core characteristics and the fuel behavior were evaluated in the analysis study at the normal operating condition. The analysis for the 9 × 9-type and the 10 × 10-type fuel assemblies and the reactor type of the Advanced BWR (ABWR) revealed a good applicability of FeCrAl-ODS fuel cladding. Finally, the challenges and perspectives found in the program are highlighted to enhance international collaborations to promote the development of the FeCrAl-ODS fuel cladding.

ジャーナルJournal of Nuclear Materials
出版ステータスPublished - 2021 12月 15

ASJC Scopus subject areas

  • 核物理学および高エネルギー物理学
  • 材料科学(全般)
  • 原子力エネルギーおよび原子力工学


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