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Title: Heterogeneous dislocation loop formation near grain boundaries in a neutron-irradiated commercial FeCrAl alloy

FeCrAl alloys are an attractive materials class for nuclear power applications due to their increased environmental compatibility over more traditional nuclear materials. Preliminary studies into the radiation tolerance of FeCrAl alloys under accelerated neutron testing between 300-400 °C have shown post-irradiation microstructures containing dislocation loops and Cr-rich ' phase. Although these initial works established the post-irradiation microstructures, little to no focus was applied towards the influence of pre-irradiation microstructures on this response. Here, a well annealed commercial FeCrAl alloy, Alkrothal 720, was neutron irradiated to 1.8 dpa at 382 °C and then the role of random high angle grain boundaries on the spatial distribution and size of dislocation loops, dislocation loops, and black dot damage was analyzed using on-zone scanning transmission electron microscopy. Results showed a clear heterogeneous dislocation loop formation with dislocation loops showing an increased number density and size, black dot damage showing a significant number density decrease, and an increased size of dislocation loops in the vicinity directly adjacent to the grain boundary. Lastly, these results suggest the importance of the pre-irradiation microstructure on the radiation tolerance of FeCrAl alloys.
ORCiD logo [1] ; ORCiD logo [2] ;  [1] ;  [1] ; ORCiD logo [1] ;  [2]
  1. Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)
  2. Univ. of Wisconsin-Madison, Madison, WI (United States)
Publication Date:
Grant/Contract Number:
Accepted Manuscript
Journal Name:
Journal of Nuclear Materials
Additional Journal Information:
Journal Volume: 483; Journal Issue: C; Journal ID: ISSN 0022-3115
Research Org:
Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States). High Flux Isotope Reactor (HFIR)
Sponsoring Org:
USDOE Office of Nuclear Energy (NE); USDOE Office of Science (SC), Basic Energy Sciences (BES) (SC-22)
Country of Publication:
United States
11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS; 36 MATERIALS SCIENCE; FeCrAl; accident tolerant; phase stability; dislocation; grain boundary
OSTI Identifier:
Alternate Identifier(s):
OSTI ID: 1414020