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Report No.

Fuel temperature analyses of metallic fuel pins for sodium-cooled fast reactors

Mizuno, Tomoyasu; Koyama, Shinichi; Kaito, Takeji; Uwaba, Tomoyuki; Tanaka, Kenya

Metallic fuel, U-Pu(TRU)-Zr is a fuel candidate for Sodium-cooled fast reactor (SFR) selected as a possible promising future nuclear reactor system in Generation-IV international forum (GIF). Design studies were performed in the Japanese feasibility study on commercialized fast reactor cycle system, and the irradiation behavior of metallic fuel is under investigation through analytical fuel performance code calculations with preliminary analytical models. Some calculations of U-Pu(TRU)-Zr fuel irradiation performance were conducted by a simplified calculation grogram developed in JAEA. Axial profile of fuel pin centerline temperature calculated by using effective fuel thermal conductivity where sodium ingress into fuel was considered fits well with actual fuel micro structures after the irradiation. The effective fuel thermal conductivity with sodium ingress is suitable for the irradiation behavior investigation.



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