Thermochemical Modeling of the Uranium-Cerium-Oxygen System
- ORNL
The objective of the Fuel Cycle R&D Program, Advanced Fuels campaign is to provide the research and development necessary to develop low loss, high quality nuclear fuels for ultra-high burnup reactor operation. Primary work in this area will be focused on the ceramic and metallic fuel systems. The goal of the current work is to enhance the understanding of ceramic nuclear fuel thermochemistry to support fuel research and development efforts. The thermochemical behavior of oxide nuclear fuel under irradiation is dependent on the oxygen to metal ratio (O:M). In fluorite-structured fuel, the actinide metal cation is bonded with {approx}2 oxygen atoms on a crystal lattice and as the metal atoms fission, fission fragments and free oxygen are created. The resulting fission fragments will contain some oxide forming elements, however these are insufficient to bind to all the liberated oxygen and therefore, there is an average increase in O:M with fuel burnup. Some of the fission products also form species that will migrate to and react with the cladding surface in a phenomenon known as Fuel Clad Chemical Interaction (FCCI). Cladding corrosion is life-limiting so it is desirable to understand influencing factors, such as oxide thermochemistry, which can be used to guide the design and fabrication of higher burn up fuel. A phased oxide fuel thermochemical model development effort is underway within the Advanced Fuels Campaign. First models of binary oxide systems are developed. For nuclear fuel system this means U-O and transuranic systems such as Pu-O, Np-O and Am-O. Next, the binary systems will be combined to form pseudobinary systems such as U-Pu-O, etc. The model development effort requires the use of data to allow optimization based on known thermochemical parameters as a function of composition and temperature. Available data is mined from the literature and supplemented by experimental work as needed. Due to the difficulty of performing fuel fabrication development with actinide materials, fundamental studies with uranium are performed using surrogate materials as stand-ins for transuranic elements. In most cases, cerium can be used as a suitable substitute for plutonium when performing O:M and sintering kinetics studies because of identical valence states. Differences exist between the magnitude of reported thermodynamic data of (U,Pu)O{sub x} and (U,Ce)O{sub x}, however the change in oxygen potential versus O:M follows the same trend for both systems. Cerium is also a major fission product element, and thus understanding its behavior in fuel is an important issue as well.
- Research Organization:
- Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)
- Sponsoring Organization:
- USDOE Office of Nuclear Energy (NE)
- DOE Contract Number:
- DE-AC05-00OR22725
- OSTI ID:
- 1004440
- Report Number(s):
- ORNL/TM-2010/214; AF5810000; NEAF224; TRN: US1100973
- Country of Publication:
- United States
- Language:
- English
Similar Records
U.S. Plans for the Next Fast Reactor Transmutation Fuels Irradiation Test
FCRD Advanced Reactor (Transmutation) Fuels Handbook
Related Subjects
21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS
ACTINIDES
ATOMS
CERAMICS
CERIUM
CRYSTAL LATTICES
FABRICATION
FISSION
FISSION FRAGMENTS
FISSION PRODUCTS
FUEL CYCLE
FUEL SYSTEMS
IRRADIATION
KINETICS
NUCLEAR FUELS
OPTIMIZATION
OXIDES
OXYGEN
OXYGEN POTENTIAL
PLUTONIUM
REACTOR OPERATION
THERMODYNAMICS
URANIUM
thermochemistry
oxygen potential
urania-ceria