2023 Annual Meeting

Presentation information

Oral presentation

V. Nuclear Fuel Cycle and Nuclear Materials » 501-2 Nuclear Fuel and the Irradiation Behavior

[2F11-15] Next Generation Fuel, Accident Tolerant Fuel

Tue. Mar 14, 2023 3:35 PM - 5:00 PM Room F (12 Bildg.3F 1232)

Chair:Shun Hirooka(JAEA)

4:35 PM - 4:50 PM

[2F15] Behavior of Chromium-coated Zirconium alloy fuel rods in simulated beyond design basis accident

*Kinya Nakamura1, Kenta Inagaki1, Juri Stuckert2, Martin Ševeček3 (1. CRIEPI, 2. KIT, 3. CTU)

Keywords:accident tolerant fuel, chromium coating, zirconium alloy, beyond design basis accident, fuel bundle

Single rod and 3×3 type bundle with Cr-coated Zr-based claddings were tested under simulated beyond design-basis accident conditions in high temperature steam environment with a peak cladding surface temperature up to 1600°C. Characterization results such as the microstructure formed between the Cr coating layer and the Zr-based alloy substrate are reported, as well as the ballooning and bursting behavior.