Small-scale mechanical testing and characterization of fuel cladding chemical interaction between HT9 cladding and advanced U-based metallic fuel alloy

Wang, Yachun; Frazer, David M.; Cappia, Fabiola; Teng, Fei; Murray, Daniel J.; Yao, Tiankai; Judge, Colin D.; Harp, Jason M.; Capriotti, Luca

doi:10.1016/j.jnucmat.2022.153754

Small-scale mechanical testing and characterization of fuel cladding chemical interaction between HT9 cladding and advanced U-based metallic fuel alloy

Journal Article · Tue May 03 00:00:00 EDT 2022 · Journal of Nuclear Materials

DOI:https://doi.org/10.1016/j.jnucmat.2022.153754· OSTI ID:1903734

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Idaho National Lab. (INL), Idaho Falls, ID (United States)
Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

Fuel cladding chemical interactions (FCCI) occurred on the interface between the nuclear metal fuel and cladding is the primary cause of cladding wastage, weakening cladding mechanical integrity, and placing fuel and cladding at risk. Although the microstructural and phase information of FCCI has been fairly understood, mechanical properties remain less studied due to limited reaction volume. Here, through a combining of advanced electron microscopy characterizations and small-scale mechanical testing techniques, including indentation and micro-tensile testing, this study investigated the microscale mechanical properties of FCCI between the ferritic/martensitic (F/M) HT9 cladding and an advanced Uranium (U)-based metallic fuel irradiated at the Advanced Test Reactor to 2.2% FIMA with peak inner cladding temperature reached to 650 °C. Mechanical testing results show significant hardening and embrittlement in the FCCI region. The brittle fracture of FCCI specimen is mainly attributed to the formation of nano-crystallized intermetallic σ-FeCr phase. Whereas mechanical softening was revealed in the unreacted HT9 matrix due to irradiation-induced microstructural and microchemical evolution, specifically, the disappearance of martensitic lath structure and the formation of Fe₂Mo Laves phase precipitation which consumed the solid solution strengthening Mo from the F/M HT9 matrix. Due to the achieved high cladding temperature, this fuel pin is of particular significance for revealing the high-temperature irradiation effect on the mechanical properties of HT9 cladding. Therefore, the outcomes of this study are expected to contribute to the development of multi-scale mechanical behavior modeling of HT9 cladding for Generation IV reactors which requires cladding to run at higher temperature (above 600 ?).

View Accepted Manuscript (DOE)

Research Organization:: Idaho National Laboratory (INL), Idaho Falls, ID (United States)

Sponsoring Organization:: USDOE Laboratory Directed Research and Development (LDRD) Program; USDOE Office of Nuclear Energy (NE)

Grant/Contract Number:: AC07-05ID14517

OSTI ID:: 1903734

Alternate ID(s):: OSTI ID: 1961031

Report Number(s):: INL/JOU-22-65574-Rev000

Journal Information:: Journal of Nuclear Materials, Journal Name: Journal of Nuclear Materials Vol. 566; ISSN 0022-3115

Publisher:: ElsevierCopyright Statement

Country of Publication:: United States

Language:: English

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