Correlation between Grain Size Distribution and Grain Boundary Character in Polycrystalline Uranium Dioxide

Journal Article · Sat Jul 01 00:00:00 EDT 2017 · Transactions of the American Nuclear Society

OSTI ID:23047464

Rudman, K.; Shaffer, B.; Lim, H.; Mcdonald, R.; Peralta, P. ^[1]

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1. School for Engineering of Matter, Transport, and Energy Arizona State University 501 E Tyler Mall, Tempe, AZ 85281 (United States)

Thermo-mechanical behavior of oxide nuclear fuels is extremely important to overall fuel element performance. Critical to the understanding of Pellet-Cladding Mechanical Interactions (PCMI) are elasticity, plasticity and creep, as this interaction with the cladding can lead to fuel rod fracture and in turn affect its thermal performance significantly. In addition, sub-grain scale mechanical behavior, e.g. elastic anisotropy, as well as dislocation-driven plasticity and creep, can also play a significant role on microstructure evolution of the fuel elements and their cladding, failure of which would severely reduce the useful service life of the elements and allow the release of fission products out into the working fluid of the reactor. Therefore, careful measurements of mechanical properties are key to validate robust fuel performance codes able to predict this PCMI behavior from inputs at the micro-scale, e.g., MARMOT, and to quantify its effects on other aspects of fuel behavior. In uranium dioxide (UO{sub 2}), the thermomechanical response at the meso-scale depends strongly on crystallography of individual grains, and available datasets are far from complete in terms of temperature, stress, and stoichiometry in parameter space. Therefore, details of the fuel microstructure, such as porosity distribution and location, as well as grain boundary (GB) character, distribution, and location are needed to create accurate multiscale models to predict the fuel's evolution in-pile, and in turn aid in the design and development of more efficient fuels that can last longer cycles.

• OSTI ID: 23047464
• Journal Information: Transactions of the American Nuclear Society, Vol. 116; Conference: 2017 Annual Meeting of the American Nuclear Society, San Francisco, CA (United States), 11-15 Jun 2017; Other Information: Country of input: France; 7 refs.; available from American Nuclear Society - ANS, 555 North Kensington Avenue, La Grange Park, IL 60526 (US); ISSN 0003-018X
• Country of Publication: United States
• Language: English

Related Subjects

36 MATERIALS SCIENCE
11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS
ANISOTROPY
CLADDING
CREEP
CRYSTALLOGRAPHY
DESIGN
ELASTICITY
FISSION PRODUCT RELEASE
FRACTURES
FUEL PELLETS
FUEL RODS
GRAIN BOUNDARIES
GRAIN SIZE
NUCLEAR FUELS
PERFORMANCE
PLASTICITY
POLYCRYSTALS
POROSITY
STOICHIOMETRY
URANIUM DIOXIDE
WORKING FLUIDS