INITIAL RESULTS FROM SAFETY TESTING OF US AGR-2 IRRADIATION TEST FUEL

Conference · Tue Nov 01 00:00:00 EDT 2016

DOI:https://doi.org/10.1016/j.nucengdes.2017.08.006· OSTI ID:1408624

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

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1. ORNL
2. Idaho National Laboratory (INL)

Two cylindrical compacts containing tristructural isotropic (TRISO)-coated particles with kernels that contained both 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 release 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 similar to the release from the UCO compacts in that an initial release of the majority of silver trapped in the matrix was observed to occur 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 may 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
• DOE Contract Number: AC05-00OR22725
• OSTI ID: 1408624
• Resource Relation: Journal Volume: 329; Conference: Winter ANS Meeting HTR 2016 - Reno, Nevada, United States of America - 11/6/2016 12:00:00 AM-11/10/2016 12:00:00 AM
• Country of Publication: United States
• Language: English