Scanning electron microscopy examination of a Fast Flux Test Facility irradiated U-10Zr fuel cross section clad with HT-9
Observations from a scanning electron microscopy examination of irradiated U-10Zr fuel are presented. The sample studied had a local burnup of 5.7 atom percent and a local inner cladding temperature of 615°C. This examination by electron microscopy has concentrated on producing data relevant to facilitating a better understanding of Zr redistribution in irradiated U-10Zr fuel and on a better understanding of the complex microstructure present in fuel cladding chemical interaction (FCCI) layers. The presented zirconium redistribution data supplements the existing literature by providing a data set at these particular local conditions. In addition to FCCI layers that are readily visible in optical microscopy, this examination has revealed lanthanide degradation of the cladding by what appears to be a grain boundary facilitated pathway. Precipitates of fission produced Pd-lanthanide compounds were observed in the fuel. Furthermore, precipitated regions with elevated Mo and elevated W content were also observed in the HT-9 cladding of this sample.
- Research Organization:
- Idaho National Lab. (INL), Idaho Falls, ID (United States)
- Sponsoring Organization:
- USDOE Office of Nuclear Energy (NE), Fuel Cycle Technologies (NE-5); USDOE Office of Nuclear Energy (NE)
- Grant/Contract Number:
- AC07-05ID14517
- OSTI ID:
- 1481288
- Alternate ID(s):
- OSTI ID: 1495575
- Report Number(s):
- INL/JOU-17-42679-Rev000
- Journal Information:
- Journal of Nuclear Materials, Vol. 494, Issue C; ISSN 0022-3115
- Publisher:
- ElsevierCopyright Statement
- Country of Publication:
- United States
- Language:
- English
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