Analysis of radially resolved thermal conductivity in high burnup mixed oxide fuel and comparison to thermal conductivity correlations implemented in fuel performance codes

Ferrigno, Joshua; Pavlov, Tsvetoslav; Poudel, Narayan; Salvato, Daniele; Tsai, Chuting Tan; Merritt, Brian; Hansen, Alex; Munro, Troy; Cappia, Fabiola; Khafizov, Marat

doi:10.1016/j.jnucmat.2024.155090

Title: Analysis of radially resolved thermal conductivity in high burnup mixed oxide fuel and comparison to thermal conductivity correlations implemented in fuel performance codes

Journal Article · 14 April 2024 · Journal of Nuclear Materials

DOI: https://doi.org/10.1016/j.jnucmat.2024.155090 · OSTI ID:2372622

^[1]; Pavlov, Tsvetoslav ^[2];

^[2];

^[3]; Hansen, Alex ^[3]; Munro, Troy ^[3];

^[2];

^[1]

The Ohio State Univ., Columbus, OH (United States)
Idaho National Laboratory (INL), Idaho Falls, ID (United States)
Brigham Young Univ., Provo, UT (United States)

The thermal diffusivity and thermal conductivity of high burnup (19 % FIMA) mixed oxide (U, Pu)O₂ nuclear fuel has been measured along the radial direction using a thermoreflectance-based method. Measured thermal conductivity exhibits a notable radial variation consistent with the expectations that a large temperature gradient across the annular fuel pellet leads to a heterogeneous microstructure. A common fuel performance model of thermal conductivity, the Lucuta-Inoue model, is used to analyze the measured thermal conductivity profile. Further, this model adequately captures the radial dependence of thermal conductivity except in the periphery. The analysis suggests that the characteristic radial shape of the thermal conductivity profile follows the burnup profile within the fuel pin. In the periphery, the high burnup structure is formed and the conductivity model, not capturing this effect, likely overestimates the thermal conductivity.

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Research Organization:: Idaho National Laboratory (INL), Idaho Falls, ID (United States)

Sponsoring Organization:: USDOE Laboratory Directed Research and Development (LDRD) Program; USDOE Office of Nuclear Energy (NE)

Grant/Contract Number:: AC07-05ID14517

OSTI ID:: 2372622

Report Number(s):: INL/JOU--23-72391-Rev000

Journal Information:: Journal of Nuclear Materials, Journal Name: Journal of Nuclear Materials Vol. 596; ISSN 0022-3115

Publisher:: ElsevierCopyright Statement

Country of Publication:: United States

Language:: English

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Title: Analysis of radially resolved thermal conductivity in high burnup mixed oxide fuel and comparison to thermal conductivity correlations implemented in fuel performance codes

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