Corrosion of 316H stainless steel in flowing FLiNaK salt

Raiman, Stephen S.; Kurley III, J. Matthew; Sulejmanovic, Dino; Willoughby, Adam W.; Nelson, Scott W.; Mao, Keyou; Parish, Chad M.; Greenwood, Michael Scott; Pint, Bruce A.

doi:10.1016/j.jnucmat.2022.153551

Title: Corrosion of 316H stainless steel in flowing FLiNaK salt

Journal Article · 19 January 2022 · Journal of Nuclear Materials

DOI: https://doi.org/10.1016/j.jnucmat.2022.153551 · OSTI ID:1846548

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^[2]; Sulejmanovic, Dino ^[2]; Willoughby, Adam W. ^[2]; Nelson, Scott W. ^[2];

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Texas A & M Univ., College Station, TX (United States)
Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

In this report type 316H stainless steel samples were exposed to flowing FLiNaK salt for 1000 h in a monometallic thermal convection loop (TCL) with a maximum temperature of 650 °C and a minimum of 540 °C. Samples in the hottest part of the TCL lost mass, with a maximum mass loss of 1.4 mg/cm², while samples in the coldest parts of the TCL gained mass, with a maximum mass gain of 1.0 mg/cm². Analysis of the samples that gained mass showed an Fe-rich layer on the sample surfaces, indicating that Fe, not Cr, was the primary deposition product in the TCL. Cr loss was apparent to a depth of ~5 µm in the hot leg. Post-exposure analysis of the salt showed major increases in the Cr, Fe, and Mn contents. The TCL was modeled using the TRANSFORM code. Modeled values matched the experimental temperature measurements showing that TRANSFORM is capable of accurately simulating the TCL conditions.

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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:: 1846548

Journal Information:: Journal of Nuclear Materials, Journal Name: Journal of Nuclear Materials Journal Issue: 4 Vol. 561; ISSN 0022-3115

Publisher:: ElsevierCopyright Statement

Country of Publication:: United States

Language:: English

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