b'Fully Ceramic Microencapsulated Fuel Pellet Production for Irradiation in the High Flux Isotope ReactorPrincipal Investigator: Rachel L. Seibert, ORNLTeam Members/ Collaborators: Jim O. Kiggans Jr., Kurt A. Terrani, and Joseph R. Burns (ORNL)The fully ceramicF an approximate diameter of 5.8 mm ully ceramic microencapsulated (FCM) fuel is an accident-tolerantand 2.8 mm height. Uranium nitride microencapsulated fuelconcept under considerationand two-phase uranium carbide/development project hasfor use in light water and advanceduranium oxide TRISO coated fuel conducted pre-irradiationnuclear reactors. FCM fuel utilizes thekernels were overcoated with a characterization and hastristructural-isotropic (TRISO) fuellow-viscosity silicon carbide mixture supplied fuel samplesdesign, embedded within a siliconusing a custom-built rotating chamber for irradiation testingcarbide (SiC) matrix. This design isrotated at 80 rpm and heated to 80C. in the high flux isotopeinherently safe with the ability toThe coating was intended to aid in reactor in FY20 to furtherachieve higher burnup, improvedfuel particle compaction and densifica-the development of nextthermal performance, and improvedtion during sintering. These coated generation nuclear fuelreliability tailored to the fuel kernelparticles were loaded in the middle under the AFC campaign. and targeted reactor type. The workof silicon carbide nanopowder with here involved the fabrication of FCMyttria-alumina sintering aides into a fuel pellets for irradiation studies in the5.8 mm graphite die. They were hot high flux isotope (HFIR) reactor at Oakpressed at 1,800C for 30 minutes Ridge National Laboratory (ORNL).with a pressure of approximately 9.5 These pellets utilize the miniature fuelMPa to densify. Neutronics calculations concept, which is intended to morewere performed at positions closest to rapidly examine and qualify differentthe HFIR cores midplane to determine fuel forms due to small sample size andmaximum neutron and gamma heating rapid burnup accumulation. Uraniumrates for the fuel. This information is nitride and two-phase uraniumused to predict fuel burnup rate and for carbide/uranium oxide kernels werefuture thermal analysis. They are carried considered and will be evaluatedout using Monte Carlo N-Particle 5 after irradiation for performance and(MCNP5) static transport calculations integrity of FCM fuels under variousand COUPLE and ORIGIN modules reactor conditions. of the SCALE code for depletion and Project Description: decay. The calculations for fuel deple-FCM pellets were fabricated to fit ation assumed fixed conditions for flux, modified version of the Miniaturespectrum, and reaction cross-sections. Fuel irradiation capsules designed byNo degradation of coated particles was Petrie et al. at ORNL. They maintainobserved during sample compaction, and an average compact density of 97% was obtained. 68 2019|AFC ACCOMPLISHMENTS'