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

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

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

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

Research output: Contribution to journal › Article › peer-review

4 Scopus citations

Abstract

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. 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.

Original language	English
Article number	154419
Journal	Journal of Nuclear Materials
Volume	580
DOIs	https://doi.org/10.1016/j.jnucmat.2023.154419
State	Published - Jul 2023

Funding

Notice: This manuscript has been authored by UT-Battelle, LLC, under contract DE-AC05–00OR22725 with the US Department of Energy (DOE). The US government retains and the publisher, by accepting the article for publication, acknowledges that the US government retains a nonexclusive, paid-up, irrevocable, worldwide license to publish or reproduce the published form of this manuscript, or allow others to do so, for US government purposes. DOE 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 ). This work was supported by the Transformational Challenge Reactor Program of the US Department of Energy , Office of Nuclear Energy . Michael Trammell and Brian Jolly (formerly at ORNL now at Ultra Safe Nuclear Corporation) contributed to the original development and fabrication of the TCR fuel form. Many other ORNL staff members contributed to this work, including Dylan Richardson (binder jet printing), Zach Burns and Tyson Jordan (post-irradiation examination), Becky Johnson (radiological control support), and Chris Hobbs (pre-irradiation characterization). John Hunn (ORNL) provided insights in interpreting the experimental results. Ian Greenquist (ORNL) and Mitchell Meyer (Ultra Safe Nuclear Corporation) provided helpful comments on the manuscript. Nesrin Cetiner, Gyutae Park, and Mike Ames (MIT) contributed to the calculations of fuel power density and fission product activities. This work was supported by the Transformational Challenge Reactor Program of the US Department of Energy, Office of Nuclear Energy. Michael Trammell and Brian Jolly (formerly at ORNL now at Ultra Safe Nuclear Corporation) contributed to the original development and fabrication of the TCR fuel form. Many other ORNL staff members contributed to this work, including Dylan Richardson (binder jet printing), Zach Burns and Tyson Jordan (post-irradiation examination), Becky Johnson (radiological control support), and Chris Hobbs (pre-irradiation characterization). John Hunn (ORNL) provided insights in interpreting the experimental results. Ian Greenquist (ORNL) and Mitchell Meyer (Ultra Safe Nuclear Corporation) provided helpful comments on the manuscript. Nesrin Cetiner, Gyutae Park, and Mike Ames (MIT) contributed to the calculations of fuel power density and fission product activities.

Keywords

Additive manufacturing
Fission gas release
Silicon carbide
Tristructural-isotropic
Uranium nitride

UN SDGs

This output contributes to the following UN Sustainable Development Goals (SDGs)

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10.1016/j.jnucmat.2023.154419

Cite this

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

abstract = "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. 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.",

keywords = "Additive manufacturing, Fission gas release, Silicon carbide, Tristructural-isotropic, Uranium nitride",

author = "Petrie, {Christian M.} and Linton, {Kory D.} and Gokul Vasudevamurthy and Danny Schappel and Seibert, {Rachel L.} and David Carpenter and Nelson, {Andrew T.} and Terrani, {Kurt A.}",

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AU - Petrie, Christian M.

AU - Linton, Kory D.

AU - Vasudevamurthy, Gokul

AU - Schappel, Danny

AU - Seibert, Rachel L.

AU - Carpenter, David

AU - Nelson, Andrew T.

AU - Terrani, Kurt A.

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AB - 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. 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.

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

Abstract

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Keywords

UN SDGs

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