UC-ZrC (NbC) as a Reactor Fuel Material

• Published in: Journal of the Atomic Energy Society of Japan / Atomic Energy Society of Japan Vol. 6; no. 4; pp. 197 - 205
• Main Author: YAMAMOTO, Kazuo
• Format: Journal Article
• Language: Japanese
• Published: Atomic Energy Society of Japan 1964
• ISSN: 0004-7120; 2186-5256
• DOI: 10.3327/jaesj.6.197

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.