Compatibility of Zircaloy-4 in eutectic PbSn and PbBiSn liquid metals

Marie Romedenne; Caleb Massey; Ercan Cakmak; Ian Greenquist

doi:10.1016/j.jnucmat.2025.155911

Compatibility of Zircaloy-4 in eutectic PbSn and PbBiSn liquid metals

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Abstract

A novel nuclear fuel concept has been proposed, consisting of UO₂ particles suspended in a liquid metal alloy. To evaluate its feasibility, the compatibility of Zircaloy-4 cladding with candidate eutectic PbSn and PbBiSn alloys—as well as with pure Pb, Bi, and Sn—was investigated through 1000-hour exposures at 400 °C and 600 °C. Severe degradation of the Zircaloy-4 specimens was observed in eutectic PbSn at 600 °C, whereas oxidation and oxygen diffusion were evident after exposure at 400 °C. At 600 °C, the formation of Zr₃Fe precipitates at the oxide–alloy interface was identified in specimens exposed to Pb and PbBiSn. Thermodynamic analysis suggested that the observed intermetallic formation resulted from the destabilization of Laves phases due to oxygen ingress at elevated temperatures.

Original language	English
Article number	155911
Journal	Journal of Nuclear Materials
Volume	615
DOIs	https://doi.org/10.1016/j.jnucmat.2025.155911
State	Published - Sep 2025

Funding

The authors gratefully acknowledge M. Stephens, B. Johnston, A. Willoughby, C. De Lamater-Brotherton, D. Sulejmanovic, T. Lowe, D. Newberry, and J. Horenburg for their valuable assistance with the experimental work. The authors also thank the Nuclear Nonproliferation Division for providing access to the Zeiss Gemini 460 SEM. This research was supported by the Laboratory Directed Research and Development (LDRD) SEED funding program. This manuscript has been authored by UT-Battelle, LLC under Contract No. DE-AC05-00OR22725 with the U.S. Department of Energy. The United States Government retains and the publisher, by accepting the article for publication, acknowledges that the United States Government retains a non-exclusive, paid-up, irrevocable, world-wide license to publish or reproduce the published form of this manuscript, or allow others to do so, for United States Government purposes. The Department of Energy will provide public access to these results of federally sponsored research in accordance with the DOE Public Access Plan ( http://energy.gov/downloads/doe-public-access-plan ).

Keywords

Bismuth
Cladding
Corrosion
Lead
Tin
Zircaloy-4

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10.1016/j.jnucmat.2025.155911

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title = "Compatibility of Zircaloy-4 in eutectic PbSn and PbBiSn liquid metals",

abstract = "A novel nuclear fuel concept has been proposed, consisting of UO₂ particles suspended in a liquid metal alloy. To evaluate its feasibility, the compatibility of Zircaloy-4 cladding with candidate eutectic PbSn and PbBiSn alloys—as well as with pure Pb, Bi, and Sn—was investigated through 1000-hour exposures at 400 °C and 600 °C. Severe degradation of the Zircaloy-4 specimens was observed in eutectic PbSn at 600 °C, whereas oxidation and oxygen diffusion were evident after exposure at 400 °C. At 600 °C, the formation of Zr₃Fe precipitates at the oxide–alloy interface was identified in specimens exposed to Pb and PbBiSn. Thermodynamic analysis suggested that the observed intermetallic formation resulted from the destabilization of Laves phases due to oxygen ingress at elevated temperatures.",

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doi = "10.1016/j.jnucmat.2025.155911",

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T1 - Compatibility of Zircaloy-4 in eutectic PbSn and PbBiSn liquid metals

AU - Romedenne, Marie

AU - Massey, Caleb

AU - Cakmak, Ercan

AU - Greenquist, Ian

PY - 2025/9

Y1 - 2025/9

N2 - A novel nuclear fuel concept has been proposed, consisting of UO₂ particles suspended in a liquid metal alloy. To evaluate its feasibility, the compatibility of Zircaloy-4 cladding with candidate eutectic PbSn and PbBiSn alloys—as well as with pure Pb, Bi, and Sn—was investigated through 1000-hour exposures at 400 °C and 600 °C. Severe degradation of the Zircaloy-4 specimens was observed in eutectic PbSn at 600 °C, whereas oxidation and oxygen diffusion were evident after exposure at 400 °C. At 600 °C, the formation of Zr₃Fe precipitates at the oxide–alloy interface was identified in specimens exposed to Pb and PbBiSn. Thermodynamic analysis suggested that the observed intermetallic formation resulted from the destabilization of Laves phases due to oxygen ingress at elevated temperatures.

AB - A novel nuclear fuel concept has been proposed, consisting of UO₂ particles suspended in a liquid metal alloy. To evaluate its feasibility, the compatibility of Zircaloy-4 cladding with candidate eutectic PbSn and PbBiSn alloys—as well as with pure Pb, Bi, and Sn—was investigated through 1000-hour exposures at 400 °C and 600 °C. Severe degradation of the Zircaloy-4 specimens was observed in eutectic PbSn at 600 °C, whereas oxidation and oxygen diffusion were evident after exposure at 400 °C. At 600 °C, the formation of Zr₃Fe precipitates at the oxide–alloy interface was identified in specimens exposed to Pb and PbBiSn. Thermodynamic analysis suggested that the observed intermetallic formation resulted from the destabilization of Laves phases due to oxygen ingress at elevated temperatures.

KW - Bismuth

KW - Cladding

KW - Corrosion

KW - Lead

KW - Tin

KW - Zircaloy-4

UR - https://www.scopus.com/pages/publications/105007546017

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DO - 10.1016/j.jnucmat.2025.155911

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VL - 615

JO - Journal of Nuclear Materials

JF - Journal of Nuclear Materials

M1 - 155911

ER -

Compatibility of Zircaloy-4 in eutectic PbSn and PbBiSn liquid metals

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Keywords

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