|About this Abstract
||2016 TMS Annual Meeting & Exhibition
||Materials and Fuels for the Current and Advanced Nuclear Reactors V
||Atomic-level Characterization of the Metal-oxide Interface of a Zircaloy-4 Cladding from Commercial LWR Irradiated Fuel
||Philip Edmondson, Chad M Parish, Tyler J Gerczak, Keith J Leonard, Arthur T Motta, Kurt A Terrani
|On-Site Speaker (Planned)
Zirconium-based alloys are the standard cladding material for nuclear fuel in light water reactors. Minor alloying elements (i.e. Sn, Fe, Cr, or Nb) greatly reduce waterside corrosion of these alloys under normal operating conditions, which may be 4-5 years and generate oxide layers of 20 to 150 micrometers. In addition to alloy composition, the oxidation behavior of these alloys is a strong function of the microstructure of both the metal and the growing oxide. In this work, we will present the results of atom probe tomography and analytical transmission electron microscopy at the metal/oxide interface to examine these multi-synergistic effects in a Zircaloy-4 fuel cladding irradiated for 7 cycles in the HB Robinson nuclear power plant (Hartsville, SC, USA). These unique results are complementary to carefully conducted laboratory autoclave corrosion tests that are conducted in the absence of neutron/gamma irradiation.
||Planned: A print-only volume