Burst and oxidation behavior of Cr-coated Zirlo during simulated LOCA testing

Bell, S. B.; Graening, T.; Evans, A.; Kelly, P.; Pint, B. A.; Kane, K. A.

doi:10.1016/j.jnucmat.2022.153679

Burst and oxidation behavior of Cr-coated Zirlo during simulated LOCA testing

Journal Article · Wed Jun 01 00:00:00 EDT 2022 · Journal of Nuclear Materials

DOI:https://doi.org/10.1016/j.jnucmat.2022.153679· OSTI ID:1874646

Bell, S. B. ^[1]; ^[2]; Evans, A. ^[3]; Kelly, P. ^[3]; ^[2]; Kane, K. A. ^[2]

University 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)
Manchester Metropolitan University (United Kingdom)

We report Cr-coated Zr-alloys are a near-term cladding concept to improve reactor safety during accident scenarios. Burst and steam oxidation behavior of bare and Cr-coated Zirlo claddings were examined under simulated loss-of-coolant accident conditions. The 4.4 µm coating had no substantial effect on ballooning or opening geometry but did increase burst temperatures at higher pressures. The coating reduced steam oxidation of the cladding compared to bare specimens, but in regions of high strain, the coating developed through-cracks allowing rapid underlying zirconia formation.

View Accepted Manuscript (DOE)

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

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

Grant/Contract Number:: AC05-00OR22725

OSTI ID:: 1874646

Alternate ID(s):: OSTI ID: 1961032

Journal Information:: Journal of Nuclear Materials, Journal Name: Journal of Nuclear Materials Journal Issue: 1 Vol. 564; ISSN 0022-3115

Publisher:: ElsevierCopyright Statement

Country of Publication:: United States

Language:: English

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