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Cladding inner surface wastage for mixed-oxide liquid metal reactor fuel pins

Conference ·
OSTI ID:6345371
Cladding inner surface wastage was measured on reference fuel pins with stainless steel and D9 cladding irradiated beyond goal burnup in the Fast Flux Test Facility. Measurements were compared to the Experimental Breeder Reactor No. 2 based fuel-cladding chemical interaction correlation developed for uranium-plutonium oxide fuels with 20% cold-worked stainless steel cladding. The fuel-cladding chemical interaction was also measured in fuel pins irradiated with HT9 cladding. Comparison of the measurements with the design correlation showed the correlation adequately accounted for the extent of interaction in the Fast Flux Test Facility fuel pins with cold-worked stainless steel D9, and HT9 cladding. 9 refs., 6 figs.
Research Organization:
Westinghouse Hanford Co., Richland, WA (USA)
Sponsoring Organization:
DOE/NE
DOE Contract Number:
AC06-87RL10930
OSTI ID:
6345371
Report Number(s):
WHC-SA-0944; CONF-901101--64; ON: DE91004577
Country of Publication:
United States
Language:
English

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