[1G13-18] Thermal Hydraulics in Fast Reactor

[1G16] Development of fundamental numerical simulation system for integrated safety evaluation in various innovative sodium-cooled fast reactor

(13) Model development for fuel failure determination and post-failure behavior

*Tohru Suzuki¹, Shuntaro Aono¹, Sho Fuchita², Shinya Ishida³, Yasushi Okano³ (1. TCU, 2. HGNE, 3. JAEA)

Keywords:safety evaluation, innovative sodium-cooled fast reactor, core melting, fuel failure, post-failure behavior

As a meltdown model of MOX-fueled core, an analytical model was constructed for reactivity power change, fuel failure determination, and mass transfer after failure, and its validity was confirmed and problems were identified. In addition, we extended this model to metal-fueled core and confirmed its basic functions.