Enhanced Thermal Conductivity UO{sub 2}-BeO and UO{sub 2}-SiC Fuels Behavior

Journal Article · Wed Jun 15 04:00:00 EDT 2016 · Transactions of the American Nuclear Society

OSTI ID:22992138

Liu, Rong; Zhou, Wenzhong ^[1]

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1. Department of Mechanical and Biomedical Engineering, City University of Hong Kong, Hong Kong (United States)

Commercial light-water reactors currently use UO{sub 2} as a nuclear fuel, which has demonstrated several desirable characteristics. However, UO{sub 2} has a low thermal conductivity that leads to the development of a large temperature gradient across the fuel pellet. This temperature gradient contributes to limits in the operational performance of the reactor due to effects that include thermal stresses causing pellet cladding interaction and the release of fission product gases as demonstrated. These high fuel temperatures can be decreased and reactor performance improved by developing an enhanced thermal conductivity nuclear fuel. A high thermal conductivity nuclear fuel would decrease fuel temperatures and facilitate a reduction in pellet cladding interaction through lessening thermal stresses that result in fuel cracking, relocation, and swelling as described. Additionally, fission gas release would be decreased allowing for higher fuel burn-up, and the safety of the reactor would be improved with a faster thermal response and less stored energy in the fuel pins. A ceramic-ceramic composite nuclear fuel has an increased effective thermal conductivity due to adding a second compatible and high thermal conductivity ceramic to UO{sub 2}. SiC and BeO are two materials that demonstrate compatibility with UO{sub 2} and have a high thermal conductivity. In summary, the enhanced thermal conductivity UO{sub 2}-BeO and UO{sub 2}-SiC fuel performance has been modelled in a light water reactor, and a more complete and detailed understanding was achieved for enhanced thermal conductivity UO{sub 2}-BeO and UO{sub 2}-SiC fuels. BeO and SiC have high melting point, low neutron absorption, low thermal expansion coefficient and chemical compatibility with current UO{sub 2} fuels which makes them a suitable material for thermal conductivity enhancement. In our modeling and simulation work, the enhanced thermal conductivity nuclear fuel showed the capability of decreasing the fuel temperatures, increasing the gap total conductance, reducing the fission gas release, decreasing the gap/plenum pressure, lessening oxygen redistribution, and facilitating a reduction in pellet cladding interaction through lessening thermal stresses that result in fuel cracking, relocation, and swelling. The safety of the reactor operation would be then improved. (authors)

• OSTI ID: 22992138
• Journal Information: Transactions of the American Nuclear Society, Journal Name: Transactions of the American Nuclear Society Journal Issue: 1 Vol. 114; ISSN 0003-018X
• Country of Publication: United States
• Language: English

Related Subjects

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS
BERYLLIUM OXIDES
CLADDING
FISSION PRODUCT RELEASE
FUEL PELLETS
MELTING POINTS
NUCLEAR FUELS
SILICON CARBIDES
SWELLING
THERMAL CONDUCTIVITY
THERMAL STRESSES
URANIUM DIOXIDE
WATER COOLED REACTORS
WATER MODERATED REACTORS