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- W2023824871 startingPage "197" @default.
- W2023824871 abstract "As an attempt to improve the stability problem of U-carbide fuel, the solid solutions of UC-ZrC and UC-NbC were prepared. Their stabilities against water corrosion and atmospheric oxidation and their compatibility with graphite were compared with those of UC and UC2.UO2+C+ZrC (NbC) mixture was heated in vacuum at around 1, 800°C for 1 hr to form the solid solution. By milling and compacting the reacted carbide and sintering in vacuum at 1, 900°2, 000°C for 1 hr, UC-ZrC or UC-NbC pellets having the density more than 90% of theoretical value were produced.It was revealed that UC2 tended to decompose into C and U2C3 in the temperature range lower than 1, 500°C and transformed from cubic to tetragonal lattice at about 1, 800°C, although it was compatible with graphite up to 2, 000°C. Whereas, (U. Zr)C or (U. Nb)C showed no lattice transformation up to melting point and more excellent stabilities against corrosion and oxidation, and furthermore, it was compatible with graphite up to 2, 000°C, as long as ZrC or NbC content was more than 72mol/_??_ or 63mol/_??_.These mixed carbides are promising as a fuel material for a high temperature gas-cooled reactor." @default.
- W2023824871 created "2016-06-24" @default.
- W2023824871 creator A5024204903 @default.
- W2023824871 date "1964-01-01" @default.
- W2023824871 modified "2023-10-18" @default.
- W2023824871 title "UC-ZrC (NbC) as a Reactor Fuel Material" @default.
- W2023824871 doi "https://doi.org/10.3327/jaesj.6.197" @default.
- W2023824871 hasPublicationYear "1964" @default.
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