In collaboration with industry, Idaho National Laboratory is investigating uranium silicide for use in future light water reactor fuels as a more accident resistant alternative to uranium oxide base fuels. Specifically this project was focused on producing uranium silicide (U 3 Si 2) pellets by conventional powder metallurgy with a density greater than 94% of the theoretical density. This work has produced a process to consistently produce pellets with the desired density through careful optimization of the process. Milling of the U 3 Si 2 has been optimized and high phase purity U 3 Si 2 has been successfully produced. Results are presented from sintering studies and microstructural examinations that illustrate the need for a finely ground reproducible particle size distribution in the source powder. The optimized process was used to produce pellets for the Accident Tolerant Fuel-1 irradiation experiment. The average density of these pellets was 11.54 ±0.06 g/cm 3. Additional characterization of the pellets by scaning electron microscopy and X-ray diffraction has also been performed. Pellets produced in this work have been encapsulated for irradiation, and irradiation in the Advanced Test Reactor is expected soon.

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Analysis Key information pertaining to fission product release Capsule components Cumulative inventory of fission products released from all fuel compacts in a capsule during irradiation Fuel compact gamma scanning Retained inventory of fission products in each individual fuel compact Deconsolidationleach-burn-leach Inventory of fission products retained in the compacts outside of the SiC layers Particle gamma counting Retained inventory of fission products in the particles 2.2.1. Analysis of irradiation capsule components for deposited fission products The major components of the irradiation capsules-including the metal shells and structural components, the graphite fuel holders, and graphite spacers at the ends of the capsules-were analyzed at INL to determine the cumulative inventories of fission products released from the compacts during the irradiation. As there were 12 compacts in each irradiation capsule, the data provide information on the total inventory of fission products released from all 12 fuel compacts; release from each compact cannot be derived from these data. The metal capsule shells were leached in acid to remove deposits. Analysis of the leach solutions included gamma spectrometry for gamma-emitting fission products, Sr separation and gas flow proportional counting for 90

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The Advanced Gas Reactor (AGR)-3/4 experiment was designed to study fission product transport within graphitic matrix material and nuclear-grade graphite. To this end, this experiment consisted of 12 capsules, each fueled with 4 compacts containing uranium oxycarbide (UCO) tri-structural isotropic (TRISO)-coated particles as driver fuel and 20 UCO designed-to-fail (DTF) fuel particles in each compact. The DTF fuel was fabricated with a thin pyrocarbon layer that was intended to fail during irradiation and provide a known source of fission products. These fission products could then migrate through the compact and into the surrounding concentric rings of graphitic matrix material and/or nuclear-grade graphite. Through post-irradiation examination (PIE) of the rings (including physical sampling and gamma scanning) fission product concentration profiles within the rings can be determined. These data can be used to elucidate fission product transport parameters (e.g., diffusion coefficients within the test materials) which will be used to inform and refine models of fission product transport. After irradiation in the Advanced Test Reactor (ATR) had been completed in April 2014, the AGR-3/4 experiment was shipped to the Hot Fuel Examination Facility (HFEF) at the Materials and Fuels Complex (MFC) for inspection, disassembly, and metrology. The AGR-3/4 test train was received at MFC in two separate shipments between February and April 2015. Visual examinations of the test train exterior did not indicate dimensional distortion, and only two small discolored areas were observed at the bottom of Capsules 8 and 9. Despite slight external discoloration, no corresponding discoloration was found on the inside of these capsules. Prior to disassembly, the two test train sections were subject to analysis via the Precision Gamma Scanner (PGS), which did not indicate that any gross fuel relocation had occurred. A series of specialized tools, including clamps, cutters, and drills, had been designed and fabricated to carry out test train disassembly and recovery of capsule components (graphite rings and fuel compacts). This equipment performed well for separating each capsule in the test train and extracting the capsule components. Only a few problems were encountered. In one case, the outermost ring (the sink ring) was cracked during removal of the capsule through tubes. Although the sink ring will be analyzed to obtain a mass-balance of fission products in the experiment, these cracks do not pose a major concern because the sink ring will not be analyzed for its fission product spatial distribution. In Capsules 4 and 5, the compacts could not be removed from the inner rings using standard methods. An arbor press was modified and used to successfully remove the compacts from the inner rings without damaging the rings. Dimensional measurements were made on the compacts, inner rings, outer rings, and sink rings. The diameters of all compacts decreased by 0.5 to 2.0%. Generally, the extent of compact diametric shrinkage increased with...