Wettability and CHF limits of Accident-Tolerant nuclear fuel cladding materials in light water reactor conditions

Su, G. Y.; Moreira, T. A.; Lee, D.; Jena, A.; Wang, G.; Byers, A.; Phillips, B.; Karoutas, Z.; Anderson, M.; Bucci, M.

doi:10.1016/j.applthermaleng.2022.119018

Wettability and CHF limits of Accident-Tolerant nuclear fuel cladding materials in light water reactor conditions

Journal Article · Fri Jul 22 00:00:00 EDT 2022 · Applied Thermal Engineering

DOI:https://doi.org/10.1016/j.applthermaleng.2022.119018· OSTI ID:2417770

Su, G. Y. ^[1]; Moreira, T. A. ^[2]; Lee, D. ^[2]; Jena, A. ^[3]; Wang, G. ^[4]; Byers, A. ^[4]; Phillips, B. ^[3]; Karoutas, Z. ^[4]; Anderson, M. ^[2]; Bucci, M. ^[3]

Massachusetts Inst. of Technology (MIT), Cambridge, MA (United States); OSTI
Univ. of Wisconsin, Madison, WI (United States)
Massachusetts Inst. of Technology (MIT), Cambridge, MA (United States)
Westinghouse Electric Corp., LLC, Columbia, SC (United States)

Here, we present the results of experimental investigations aimed at evaluating the thermal–hydraulic performance of chromium-coated zircaloy, i.e., one of the most promising accident tolerant fuel (ATF) cladding material for light water nuclear reactors. Precisely, we investigate the wettability and critical heat flux (CHF) limits of chromium-coated and conventional zircaloy surfaces in prototypical reactor conditions. For both surface types, we measure the contact angle in a vapor-saturated atmosphere from ambient pressure to the operating pressure of pressurized water reactors (PWRs), i.e., ~15 MPa. We measure the ambient-pressure steady-state flow boiling CHF with a spatially uniform heat flux. We measure the high-pressure steady-state flow boiling CHF with a cosine shape heat flux (up to 20 MPa) and with a uniform heat flux (up to 15 MPa), also exploring the effect of CRUD deposits on the chromium-coated surface. Our results reveal that the chromium surface and the bare zircaloy surface have similar wettability and both become super-hydrophilic in PWR conditions, and that there is practically no difference in the steady-state CHF limits, both at low-pressure and high-pressure conditions, also when the chromium-coated surface is covered by a CRUD deposit. However, while the chromium-coating does not improve the CHF compared to the bare zircaloy surface, it improves the post-CHF behavior. The chromium coating prevents the reaction between zircaloy and steam, which results in the formation of a brittle zirconium oxide through the surface of the cladding. We also measure the transient CHF under exponentially escalating heat flux inputs of a nano-smooth and a rough surface mimicking a chromium-coated zircaloy cladding. Interestingly, the results of the transient heat flux tests suggest that the CHF limit for very short periods (i.e., fast transients) is independent of the surface finish, being the same for a rough chromium surface or a nano-smooth surface.

View Accepted Manuscript (DOE)

Research Organization:: Massachusetts Inst. of Technology (MIT), Cambridge, MA (United States)

Sponsoring Organization:: USDOE Office of Nuclear Energy (NE); USDOE Office of Nuclear Energy (NE), Nuclear Energy University Program (NEUP)

Grant/Contract Number:: NE0008693

OSTI ID:: 2417770

Alternate ID(s):: OSTI ID: 1878653

Journal Information:: Applied Thermal Engineering, Journal Name: Applied Thermal Engineering Journal Issue: C Vol. 216; ISSN 1359-4311

Publisher:: ElsevierCopyright Statement

Country of Publication:: United States

Language:: English

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36 MATERIALS SCIENCE
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Wettability and CHF limits of Accident-Tolerant nuclear fuel cladding materials in light water reactor conditions

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