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This content will become publicly available on August 18, 2018

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

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 UO 2-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 UO 2 compacts exhibited cesium release from multiple particles whose SiC failed during the safety test. Europium and strontium release from these two UO 2 compacts appeared to be dominated by release from the particles with failed SiC. Silver release was characteristically like themore » 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 UO 2 TRISO with failed SiC. Failure of the SiC layer in the UO 2 fuel appears to have been dominated by CO corrosion, as opposed to the palladium degradation observed in AGR-1 UCO fuel.« less
Authors:
ORCiD logo [1] ; ORCiD logo [1] ;  [1] ; ORCiD logo [1] ; ORCiD logo [1] ;  [2]
  1. Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)
  2. Idaho National Lab. (INL), Idaho Falls, ID (United States)
Publication Date:
Grant/Contract Number:
AC05-00OR22725
Type:
Accepted Manuscript
Journal Name:
Nuclear Engineering and Design
Additional Journal Information:
Journal Volume: 329; Journal Issue: C; Journal ID: ISSN 0029-5493
Publisher:
Elsevier
Research Org:
Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)
Sponsoring Org:
USDOE Office of Nuclear Energy (NE)
Country of Publication:
United States
Language:
English
Subject:
11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS
OSTI Identifier:
1435285