Initial results from safety testing of US AGR-2 irradiation test fuel

Journal Article · Fri Aug 18 00:00:00 EDT 2017 · Nuclear Engineering and Design

DOI:https://doi.org/10.1016/j.nucengdes.2017.08.006· OSTI ID:1435285

Morris, Robert Noel ^[1]; Hunn, John D. ^[1]; Baldwin, Charles A. ^[1]; Montgomery, Fred C. ^[1]; Gerczak, Tyler J. ^[1]; Demkowicz, Paul A. ^[2]

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1. Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)
2. Idaho National Lab. (INL), Idaho Falls, ID (United States)

Two cylindrical compacts containing tristructural isotropic (TRISO)-coated particles with kernels that contained a mixture of uranium carbide and uranium oxide (UCO) and two compacts with UO₂-kernel TRISO particles have undergone 1600°C safety testing. These compacts were irradiated in the US Advanced Gas Reactor Fuel Development and Qualification Program's second irradiation test (AGR-2). The time-dependent releases of several radioisotopes (^110mAg, ¹³⁴Cs, ¹³⁷Cs, ¹⁵⁴Eu, ¹⁵⁵Eu, ⁹⁰Sr, and ⁸⁵Kr) were monitored while heating the fuel specimens to 1600°C in flowing helium for 300 h. The UCO compacts behaved similarly to previously reported 1600°C-safety-tested UCO compacts from the AGR-1 irradiation. No failed TRISO or failed SiC were detected (based on krypton and cesium release), and cesium release through intact SiC was very low. Release behavior of silver, europium, and strontium appeared to be dominated by inventory originally released through intact coating layers during irradiation but retained in the compact matrix until it was released during safety testing. Both UO₂ compacts exhibited cesium release from multiple particles whose SiC failed during the safety test. Europium and strontium release from these two UO₂ compacts appeared to be dominated by release from the particles with failed SiC. Silver release was characteristically like the release from the UCO compacts in that an initial release of the majority of silver trapped in the matrix occurred during ramping to 1600°C. However, additional silver release was observed later in the safety testing due to the UO₂ TRISO with failed SiC. Failure of the SiC layer in the UO₂ fuel appears to have been dominated by CO corrosion, as opposed to the palladium degradation observed in AGR-1 UCO fuel.

• 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: 1435285
• Alternate ID(s): OSTI ID: 1408624; OSTI ID: 1549866
• Journal Information: Nuclear Engineering and Design, Journal Name: Nuclear Engineering and Design Journal Issue: C Vol. 329; ISSN 0029-5493
• Publisher: Elsevier
• Country of Publication: United States
• Language: English

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