Strength and rupture geometry of un-irradiated C26M FeCrAl under LOCA burst testing conditions

Bell, Sam B.; Kane, Kenneth A.; Massey, Caleb P.; Pint, Bruce A.

doi:10.1016/j.jnucmat.2021.153242

Strength and rupture geometry of un-irradiated C26M FeCrAl under LOCA burst testing conditions

Journal Article · Wed Dec 01 00:00:00 EST 2021 · Journal of Nuclear Materials

DOI:https://doi.org/10.1016/j.jnucmat.2021.153242· OSTI ID:1817528

Bell, Sam B. ^[1]; Kane, Kenneth A. ^[2]; Massey, Caleb P. ^[2]; ^[2]

Univ. of Tennessee, Knoxville, TN (United States); Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)
Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

Ferritic iron-chromium-aluminum (FeCrAl) alloys are an accident tolerant fuel candidate to replace the incumbent Zr-based claddings. Nuclear grade FeCrAl alloys are marked by superior high temperature mechanical behavior and exceptional steam oxidation resistance, both of which increase safety margins during accident scenarios. In the present study, the loss of coolant accident (LOCA) burst behaviors of three un-irradiated commercially fabricated cladding materials in a simulated LOCA environment were compared: (1) T35Y2, a 1^st generation nuclear grade FeCrAl, (2) C26M, a 2^nd generation nuclear grade FeCrAl, and (3) Zircaloy-2. Both FeCrAl alloys showed improved mechanical strength and steam oxidation resistance compared to Zircaloy-2. C26M claddings burst at significantly higher temperatures for all tested engineering hoop stresses, had limited ballooning, and demonstrated preferential fuel retention behavior in terms of burst opening area and length. High temperature tensile data for C26M is also presented. For both un-irradiated FeCrAl alloys, it was found that a distinct “threshold” burst stress signified the transition between small and large openings. Finally, higher threshold hoop stresses were associated with higher uniaxial strength for the FeCrAl alloys, indicating that tensile data, rather than creep data, could be useful for predicting rupture size and assessing fuel dispersal concerns.

View Accepted Manuscript (DOE)

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

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

Grant/Contract Number:: AC05-00OR22725

OSTI ID:: 1817528

Alternate ID(s):: OSTI ID: 1813469

Journal Information:: Journal of Nuclear Materials, Journal Name: Journal of Nuclear Materials Journal Issue: 1 Vol. 557; ISSN 0022-3115

Publisher:: ElsevierCopyright Statement

Country of Publication:: United States

Language:: English

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