Ductility Evaluation of As-Hydrided and Hydride Reoriented Zircaloy-4 Cladding under Simulated Dry-Storage Condition
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Ductility Evaluation of As-Hydrided and Hydride Reoriented Zircaloy-4 Cladding under Simulated Dry-Storage Condition Y. Yan1, L. K. Plummer2, H. Ray1, T. Cook1, and H. Z. Bilheux1 1 Oak Ridge National Laboratory, Oak Ridge, TN 37831, U.S.A. 2 University of Oregon, Eugene, OR 97403, U.S.A. ABSTRACT Pre-storage drying-transfer operations and early stage storage expose cladding to higher temperatures and much higher pressure-induced tensile hoop stresses relative to normal operation in-reactor and pool storage under these conditions. Radial hydrides precipitate during cooling and could provide an additional embrittlement mechanism as the cladding temperature decreases below the ductile-to-brittle transition temperature. To simulate this behavior, unirradiated Zircaloy-4 samples were hydrided by a gas charging method to levels that encompass the range of hydrogen concentrations observed in current used fuel. Mechanical testing was carried out by the ring compression test (RCT) method at various temperatures to evaluate the sample’s ductility for both as-hydrided and post-hydride reorientation treated specimens. As-hydrided samples with higher hydrogen concentration (>800 ppm) resulted in lower strain before fracture and reduced maximum load. Increasing RCT temperatures resulted in increased ductility of the as-hydrided cladding. A systematic radial hydride treatment was conducted at various pressures and temperatures for the hydrided samples with H content around 200 ppm. Following the radial hydride treatment, RCTs on the hydride reoriented samples were conducted and exhibited lower ductility compared to as-hydrided samples. I. INTRODUCTION Hydrogen embrittlement of zirconium alloys is a growing concern in the United States due to the lack of a long-term solution for disposal of used nuclear fuel (UNF). Normal operation of nuclear fuel in a reactor results in the formation of a waterside corrosion layer and the introduction of hydrogen into the zirconium cladding. With increasing corrosion, the hydrogen concentration in the cladding will exceed its terminal solid solubility and brittle zirconium hydrides (Zr + 2H - >ZrH2) may precipitate as cladding cools, which causes cladding ductility and failure energy to decrease [1, 2]. The weakened cladding, degraded due to hydride precipitates, increases the likelihood of failure during very long-term storage and/or transportation of UNF. However, cladding residual strength is highly dependent upon the microstructural condition, especially the orientation of the zirconium-hydride precipitates. Recent research [3, 4] indicates that the precipitates can reorient to align with the radial axis under a tensile stress from the internally pressurized cladding around 150 MPa at ≈ 400°C, conditions similar to drying operations of the UNF [5, 6]. In this work, we performed controlled experiments to produce Zircaloy-4 samples with hydrides oriented in the circumferential or in the radial direction and characterized the microstructure as well as the mechanical properties of these samples.
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