Thermal mechanical assessment of a SiC-SiC-composite clad fuel pin concept in a light water reactor environment

Alvis, Jr., John Mark; Singh, Gyanender; Gamble, Kyle A.; Kamerman, David W.; Seo, Seokbin; Nantes, Katheren

doi:10.1016/j.nucengdes.2025.114458

Thermal mechanical assessment of a SiC-SiC-composite clad fuel pin concept in a light water reactor environment

Journal Article · Wed Sep 24 00:00:00 EDT 2025 · Nuclear Engineering and Design

DOI:https://doi.org/10.1016/j.nucengdes.2025.114458· OSTI ID:3011951

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Idaho National Laboratory (INL), Idaho Falls, ID (United States)
Pennsylvania State Univ., University Park, PA (United States)

Accident Tolerant Fuels (ATFs) are designed to increase coping time following an accident scenario while preserving or improving current steady state reactor operational performance. A potential ATF concept is SiC-SiC composite claddings. Fuel performance simulations were conducted on a SiC-SiC based cladding concept utilizing a multilayered approach for improved performance. This cladding concept referred to as the Duplex concept is a duplex structure composed of a monolithic SiC layer placed on the outside of a SiC-SiC composite. A liquid metal is added to fuel-cladding gap for improved heat dissipation from the fuel. The monolithic SiC layer is used to improve the coolant corrosion characteristics and protect the SiC-SiC composite layer from exposure to the coolant. The fuel performance code BISON was used to conduct fuel performance simulations on the cladding concepts. Comparisons are made with a current prototypic fuel rod design (UO₂ fuel enclosed in Zircaloy-4 cladding). Representative steady-state cases were considered for normal power and two cycle power histories. Additionally, a PCI ramp case was simulated to analyze potential anticipated operational occurrences. Transient response during a Loss of Coolant Accident and a Reactivity Initiated Accident were also simulated. This computational study demonstrated that for normal operating conditions, the SiC concept cladding performed as well as the baseline for the standard power cases evaluated. The ramping evaluations indicate potential fracturing of the SiC-SiC composite of the composite cladding compared to the Zircaloy-4 cladding due to the temperature gradient and the subsequent differential thermal conductivity degradation and swelling across the composite thickness. In conclusion, the rod fails early at low enthalpy for RIA but survives a LOCA with minimal material loss due to high temperature steam corrosion.

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

Sponsoring Organization:: USDOE Office of Nuclear Energy (NE), Nuclear Fuel Cycle and Supply Chain. Office of Advanced Fuels Technologies

Grant/Contract Number:: AC07-05ID14517

OSTI ID:: 3011951

Report Number(s):: INL/JOU--24-78239

Journal Information:: Nuclear Engineering and Design, Journal Name: Nuclear Engineering and Design Vol. 445; ISSN 0029-5493

Publisher:: ElsevierCopyright Statement

Country of Publication:: United States

Language:: English

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