b'2019 | ANNUAL REPORTmicrostructure of the PCI shows the feasibility of applying microme-chanical test methods to investigate the mechanical properties of fuel.References[1.] A.T. Motta, A. Couet, R.J. Com-stock, Corrosion of Zirconium Al-loys Used for Nuclear Fuel Clad-ding, Annu. Rev. Mater. Res. 45 (2015) 311343. doi:10.1146/annurev-matsci-070214-020951.[2.] K. Nogita, K. Une, High resolu-tion TEM of high burnup UO2 fuel, J. Nucl. Mater. 250 (1997) 244249. doi:10.1016/S0022-3115(97)00282-1[3.] T.G. Lach, D.J. Edwards, E.C. Buck, B.K. McNamara, J.M. Schwantes, R.A. Clark, Fission recoil-induced microstructural evolution of the fuel-cladding interface [FCI]in high burnup BWR fuel, J. Nucl. Mater. 521 (2019) 120125. doi:10.1016/j.jnuc-mat.2019.04.044.Figure 1. STEM-EDS results of the PCI-Distributed Partnership at a Glance UO 2fuel interface: (a) STEM bright-field image showing the ZrO 2region and the NSUF and Partners Facilities and Capabilities UO 2region; both regions contain pores. Idaho National Laboratory Irradiated Materials (b) to (i) Elemental maps of Zr, U, O, Characterization Laboratory Mo, Tc, Ru, Rh, and Pd, respectively.CollaboratorsIdaho National Laboratory Daniel Jderns (collaborator), M Nedim Cinbiz (principal investigator), Lingfeng He (collaborator),Xiang Liu (collaborator)57'