Fission gas retention of densely packed uranium carbonitride tristructural-isotropic fuel particles in a 3D printed SiC matrix

Petrie, Christian M.; Linton, Kory D.; Vasudevamurthy, Gokul; Schappel, Danny; Seibert, Rachel L.; Carpenter, David; Nelson, Andrew T.; Terrani, Kurt A.

doi:10.1016/j.jnucmat.2023.154419

Fission gas retention of densely packed uranium carbonitride tristructural-isotropic fuel particles in a 3D printed SiC matrix

Journal Article · Thu Mar 30 00:00:00 EDT 2023 · Journal of Nuclear Materials

DOI:https://doi.org/10.1016/j.jnucmat.2023.154419· OSTI ID:1968704

^[1]; Linton, Kory D. ^[1]; ^[1]; ^[1]; ^[1]; Carpenter, David ^[2]; ^[1]; ^[3]

Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)
Massachusetts Inst. of Technology (MIT), Cambridge, MA (United States)
Ultra Safe Nuclear Corporation, Oak Ridge, TN (United States)

The Transformational Challenge Reactor (TCR) fuel form was designed to contain large, densely packed uranium carbonitride (UCN) tristructural-isotropic (TRISO) fuel particles within a 3D printed SiC matrix, increasing the uranium density compared to conventional TRISO fuel forms and offering full geometric freedom for core design. Here, this work summarizes initial low-burnup, high-power irradiation testing of TCR fuel materials, including loose UCN TRISO particles and integral fuel compacts with ~55% TRISO particles by volume, to evaluate fission gas retention. Fission gasses were fully retained in all loose particle tests and in integral compacts irradiated at low (<250 °C) surface temperatures. Initial testing at higher (~700–750 °C) fuel surface temperatures showed fission gas release (FGR) and complete fracture of three compacts, but no FGR was observed in later high temperature tests (~300–750 °C) of both fueled compacts and loose TRISO particles. Calculated thermal stresses in the failed compacts were far less than the measured strength of the SiC matrix and the stresses in some failed compacts were less than those in compacts that did not show FGR. Thermal stress-induced matrix cracks also would not cause complete fracture because the tensile stresses transition to compression in the higher temperature regions. Therefore, fuel failure was likely not caused by thermal stresses and may have been related to leakage currents from the electrical heaters and erratic fuel surface temperatures that were only observed in the test for which failure was observed. In any case, the matrix cracks propagated through the coatings of TRISO particles located in the high-density matrix regions on the peripheries of the compacts, resulting in measurable fission gas release. The discussion focuses on the importance of understanding matrix density distributions and the particle-matrix interface properties to prevent matrix cracks from causing TRISO particle failures.

View Accepted Manuscript (DOE)

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:: 1968704

Alternate ID(s):: OSTI ID: 1968355

Journal Information:: Journal of Nuclear Materials, Journal Name: Journal of Nuclear Materials Vol. 580; ISSN 0022-3115

Publisher:: ElsevierCopyright Statement

Country of Publication:: United States

Language:: English

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