Mechanical properties of neutron-irradiated model and commercial FeCrAl alloys

Field, Kevin G.; Briggs, Samuel A.; Sridharan, Kumar; Howard, Richard H.; Yamamoto, Yukinori

doi:10.1016/j.jnucmat.2017.03.038

Mechanical properties of neutron-irradiated model and commercial FeCrAl alloys

Journal Article · Tue Mar 28 00:00:00 EDT 2017 · Journal of Nuclear Materials

DOI:https://doi.org/10.1016/j.jnucmat.2017.03.038· OSTI ID:1351763

^[1]; ^[2]; Sridharan, Kumar ^[2]; ^[1]; Yamamoto, Yukinori ^[1]

Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)
Univ. of Wisconsin, Madison, WI (United States)

The development and understanding of the mechanical properties of neutron-irradiated FeCrAl alloys is increasingly a critical need as these alloys continue to become more mature for nuclear reactor applications. This study focuses on the mechanical properties of model FeCrAl alloys and of a commercial FeCrAl alloy neutron-irradiated to up to 13.8 displacements per atom (dpa) at irradiation temperatures between 320 and 382 °C. Tensile tests were completed at room temperature and at 320 °C, and a subset of fractured tensile specimens was examined by scanning electron microscopy. Results showed typical radiation hardening and embrittlement indicative of high chromium ferritic alloys with strong chromium composition dependencies at lower doses. At and above 7.0 dpa, the mechanical properties saturated for both the commercial and model FeCrAl alloys, although brittle cleavage fracture was observed at the highest dose in the model FeCrAl alloy with the highest chromium content (18 wt %). Finally, the results suggest the composition and microstructure of FeCrAl alloys plays a critical role in the mechanical response of FeCrAl alloys irradiated near temperatures relevant to light water reactors.

View Accepted Manuscript (DOE)

Research Organization:: Oak Ridge National Laboratory (ORNL), Oak Ridge, TN (United States). High Flux Isotope Reactor (HFIR)

Sponsoring Organization:: USDOE Office of Nuclear Energy (NE); USDOE Office of Science (SC), Basic Energy Sciences (BES)

Grant/Contract Number:: AC05-00OR22725

OSTI ID:: 1351763

Alternate ID(s):: OSTI ID: 1419551

Journal Information:: Journal of Nuclear Materials, Journal Name: Journal of Nuclear Materials Journal Issue: C Vol. 489; ISSN 0022-3115

Publisher:: ElsevierCopyright Statement

Country of Publication:: United States

Language:: English

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Feasibility assessment of the alumina‐forming duplex stainless steels as accident tolerant fuel cladding materials for light water reactors Kim, Chaewon; Kim, Hyunmyung; Jang, Changheui International Journal of Energy Research, Vol. 44, Issue 10 https://doi.org/10.1002/er.5021	journal	December 2019

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