Thermochemical Analysis of Gas-Cooled Reactor Fuels Containing Am and Pu Oxides

Lindemer, T. B.

Oak Ridge National Laboratory

关键词: Corrosion;    Fission;    Getters;    Oxygen;    Oxides;   

DOI  :  10.2172/814208
RP-ID  :  ORNL/TM-2002/133
RP-ID  :  AC05-00OR22725
RP-ID  :  814208

美国|英语

来源: UNT Digital Library

【 摘 要 】

Literature values and estimated data for the thermodynamics of the actinide oxides and fission products are applied to explain the chemical behavior in gas-cooled-reactor fuels. Emphasis is placed on the Am-O-C and Pu-O-C systems and the data are used to plot the oxygen chemical potential versus temperature of solid-solid and solid-gas equilibria. These results help explain observations of vaporization in Am oxides, nitrides, and carbides and provide guidance for the ceramic processing of the fuels. The thermodynamic analysis is then extended to the fission product systems and the Si-C-O system. Existing data on oxygen release (primarily as CO) as a function of burnup in the thoria-urania fuel system is reviewed and compared to values calculated from thermodynamic data. The calculations of oxygen release are then extended to the plutonia and americia fuels. Use of ZrC not only as a particle coating that may be more resistant to corrosion by Pd and other noble-metal fission products, but also as a means to getter oxygen released by fission is discussed.