Destructive PIE and Safety Testing of Six AGR-5/6/7 Capsule 2 Compacts

Technical Report · Mon Sep 01 00:00:00 EDT 2025

DOI:https://doi.org/10.2172/3002189· OSTI ID:3002189

Cureton, William F. ^[1]; Gerczak, Tyler J. ^[1]; Werden, Jesse ^[1]; Helmreich, Grant W. ^[1]; Hunn, John D. ^[2]; Montgomery, Fred C. ^[1]; Jones, Matthew M. ^[1]; Hooghkirk, Martino R. ^[1]; Skitt, Darren J. ^[3]; Trewhitt, Stephen E. ^[1]; Dryman, Caleb R. ^[1]; Stempien, John D. ^[4]

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1. Oak Ridge National Laboratory (ORNL), Oak Ridge, TN (United States)
2. Ultra Safe Nuclear Corporation (USNC), Seattle, WA (United States)
3. Westinghouse Electric Corp., Pittsburgh, PA (United States)
4. Idaho National Laboratory (INL), Idaho Falls, ID (United States)

This study evaluates fission product retention and particle failure mechanisms in AGR-5/6/7 Capsule 2 uranium carbide and uranium oxide (UCO) tristructural isotropic (TRISO) fuel under high-temperature gas reactor accident-relevant conditions using high-temperature safety tests and destructive post-irradiation examination. Three Capsule 2 compacts were held isothermally at 1600°C for approximately 300 hours and one compact at 1800°C for approximately 300 hours; two additional compacts were examined in the as-irradiated state. Post-test deconsolidation–leach–burn–leach (DLBL) quantified nuclide inventories in matrix and particles. Individual particles were surveyed for radioisotope inventories, and microanalytical approaches resolved microstructural evolution and fission product distributions within the coating layers. At 1600°C, no krypton was detected above the minimum detectable limit, and cesium releases were far below a single particle equivalent, indicating the absence of full TRISO failure or SiC failures. Silver releases were limited and primarily reflected depleted post-irradiation inventories, consistent with prior compact-level exams indicating substantial in-pile ^110mAg loss. At 1800°C, cumulative ¹³⁴Cs release of approximately 2.5 particle equivalents and delayed ⁸⁵Kr totaling approximately 0.53 particle equivalents were consistent with one full TRISO failure and two SiC failures. Europium and strontium releases were roughly one order of magnitude higher than at 1600°C and comparable to AGR-1/AGR-2 high-temperature tests, with sustained late-hold rates indicating diffusion through intact coatings coupled with matrix depletion. Overall, AGR-5/6/7 Capsule 2 UCO fuel demonstrated fission product retention during safety testing consistent with prior AGR campaigns, while distinctive in-pile ^110mAg depletion and measurable 1600°C europium loss motivate targeted follow-on studies.

• Research Organization: Oak Ridge National Laboratory (ORNL), Oak Ridge, TN (United States)
• Sponsoring Organization: USDOE Office of Nuclear Energy (NE), Nuclear Fuel Cycle and Supply Chain. Advanced Fuel Campaign
• DOE Contract Number: AC05-00OR22725
• OSTI ID: 3002189
• Report Number(s): ORNL/TM--2025/4123
• Country of Publication: United States
• Language: English

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