2017 Annual Meeting

Presentation information

Oral presentation

IV. Nuclear Fuel Cycle and Nuclear Materials » 401-2 Nuclear Fuel and the Irradiation Behavior

[1I13-17] characteristics of fuel cladding

Mon. Mar 27, 2017 4:25 PM - 5:50 PM Room I (16-304 Building No.16)

Chair: Ken Kurosaki (Osaka Univ.)

4:55 PM - 5:10 PM

[1I15] The high temperature strength anisotropy of full and/or partial-recrystallized Zr-Nb alloy

*Sho Kano1, Huilong Yang1, Nagasaka Takuya2, Hiroaki Abe1 (1. The University of Tokyo, 2. National Institute for Fusion Science)

Keywords:Fuel cladding material, Zr-Nb alloy, Strength anisotropy , High temperature strength

The high temperature uniaxial tensile in full and/or partial-recrystallized Zr-Nb alloys were performed under temperature range below 973 K in order to assess the high temperature strength anisotropy of advanced fuel cladding material in light waste reactor. The strength anisotropy was decreased with increasing the testing temperature, as well as the ruptured-strain was exceeded 100 % under 973K condition without dependence on the tensile direction and microstructure.