Metallography and fuel cladding chemical interaction in fast flux test facility irradiated metallic U-10Zr MFF-3 and MFF-5 fuel pins

The Mechanistic Fuel Failure (MFF) series of metal fuel irradiations conducted in the Fast Flux Test Facility (FFTF) provides an important comparison between data generated in the Experimental Breeder Reactor (EBR-II) and that expected in a larger-scale fast reactor. The MFF fuel operated with a pea...

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Bibliographic Details
Published in:Journal of nuclear materials 2016-05, Vol.473 (C), p.167-177
Main Authors: Carmack, W.J., Chichester, H.M., Porter, D.L., Wootan, D.W.
Format: Article
Language:English
Subjects:
Breeder reactors
Cladding
FFTF
Flux
Fuel cladding chemical interaction
Fuels
Historic
Irradiated fuel pins
Irradiation
Metallography
MFF
Nuclear engineering
NUCLEAR FUEL CYCLE AND FUEL MATERIALS
Test facilities
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Summary:The Mechanistic Fuel Failure (MFF) series of metal fuel irradiations conducted in the Fast Flux Test Facility (FFTF) provides an important comparison between data generated in the Experimental Breeder Reactor (EBR-II) and that expected in a larger-scale fast reactor. The MFF fuel operated with a peak cladding temperature at the top of the fuel column, but developed peak burnup at the centerline of the core. This places the peak fuel temperature midway between the core center and the top of fuel, lower in the fuel column than in EBR-II experiments. Data from the MFF-3 and MFF-5 assemblies are most comparable to the data obtained from the EBR-II X447 experiment. The two X447 pin breaches were strongly influenced by fuel/cladding chemical interaction (FCCI) at the top of the fuel column. Post irradiation examination data from MFF-3 and MFF-5 are presented and compared to historical EBR-II data. •Irradiation and post irradiation examination of full-length metallic fast reactor fuel.•Fuel cladding chemical interaction formation in full-length metallic fast reactor fuel.•Correlation of FCCI with temperature and burnup.•Comparison of full-length reactor fuel performance with test reactor fuel performance.
ISSN:0022-3115
1873-4820
DOI:10.1016/j.jnucmat.2016.02.019