| Abstract Scope |
Zircaloy-4 possesses good radiation stability, corrosion resistance, mechanical properties, and low hydrogen uptake. Applications, however, are limited mostly to fuel cladding at operating temperatures above 300°C. Expanding the usage of zircaloy-4 has been proposed for projects including pressure vessels, test reactors, and reprocessing plants. Postweld mechanical properties, however, need to be qualified for these new operating regimes. This work investigates unwelded and TIG welded zircaloy-4 - with and without a post-weld heat treatment (PWHT) - neutron irradiated in the High Flux Isotope Reactor (target doses 1021 and 1022 n/cm2) at Oak Ridge National Laboratory. Low-temperature irradiation (60-90°C) and room-temperature tensile testing were performed. Emphasis is placed on understanding ductility as a function of weld, PWHT, and neutron dose through tensile testing and fractography. For unirradiated materials, peak total elongation (16.5%) is observed at 800°C, 1h PWHT, which also don’t exhibit undesirable ‘blocky alpha’ microstructure observed at longer (24h, 48h) PWHTs. |