Mechanical behavior of additively manufactured and wrought 316L stainless steels before and after neutron irradiation

Byun, T. S.; Garrison, Ben E.; McAlister, Michael R.; Chen, Xiang; Gussev, Maxim N.; Lach, Tim G.; Coq, Annabelle Le; Linton, Kory; Joslin, Chase B.; Carver, J. Keith; List III, Fred A.; Dehoff, Ryan R.; Terrani, K. A.

doi:10.1016/j.jnucmat.2021.152849

Title: Mechanical behavior of additively manufactured and wrought 316L stainless steels before and after neutron irradiation

Journal Article · Sat Jan 30 00:00:00 EST 2021 · Journal of Nuclear Materials

DOI:https://doi.org/10.1016/j.jnucmat.2021.152849· OSTI ID:1775219

^[1];

^[1]; McAlister, Michael R. ^[1];

^[1];

^[1]; Linton, Kory ^[1]; Joslin, Chase B. ^[1];

^[1];

^[1]; Terrani, K. A. ^[1]

Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

Fabrication of nuclear reactor components using additive manufacturing (AM) methods is now a practical option since the AM technologies have advanced to allow for building of complex parts with high quality materials. To assess the mechanical performance of printed components in reactor-relevant conditions and to build a property database for the AM 316L stainless steel (SS), mechanical testing and characterization were performed before and after neutron irradiation. In this work, miniature tensile specimens were irradiated at the High Flux Isotope Reactor (HFIR) to 0.2 and 2 displacements per atom (dpa) at 300 and 600°C. The AM 316L SS was tested in the as-built, stress-relieved, and solution-annealed conditions, and the wrought (WT) 316L SS in solution-annealed condition as a reference alloy. The baseline test result showed that the AM 316L SS, regardless of the post-build heat treatment, had higher strength than the WT 316L SS, but similar ductility. Post-irradiation tensile testing was conducted at RT, 300°C, and 500°C for selected irradiation conditions. Neutron irradiation induced significant changes in the mechanical behavior of the AM stainless steels, including both hardening and softening. Although the as-built 316L steel after 300°C irradiation showed necking just after yielding, the overall property changes of the as-printed alloy became less significant after 600°C irradiation. Irradiation-induced ductilization was also observed after the higher temperature irradiation. In general, the strength change was smaller in the relatively stronger as-built and stress-relieved AM SSs than in the solution-annealed AM and WT SSs. These relatively lower strength 316L SSs overall retained higher ductility in the irradiation conditions tested, but the stronger 316L SSs demonstrated a similar level of ductility after the higher temperature (600°C) irradiation. It is a positive assessment for the AM 316L materials that no embrittlement was observed within the test and irradiation conditions of the experiment.

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Research Organization:: Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

Sponsoring Organization:: USDOE Office of Nuclear Energy (NE)

Grant/Contract Number:: AC05-00OR22725

OSTI ID:: 1775219

Alternate ID(s):: OSTI ID: 1809455

Journal Information:: Journal of Nuclear Materials, Vol. 548; ISSN 0022-3115

Publisher:: ElsevierCopyright Statement

Country of Publication:: United States

Language:: English

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36 MATERIALS SCIENCE
Additively manufactured 316L stainless steels
Post-build heat treatment
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Title: Mechanical behavior of additively manufactured and wrought 316L stainless steels before and after neutron irradiation

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