... alloying systems of zirconium: Zircaloy-4, E635M, and E110opt. The temperature dependences of K IH and V DHC have the same behavior for all alloys but are quantitatively different. V DHC and its temperature dependence decrease for the alloys in the above sequence, whereas K IH increases...

... at temperatures found during interim dry storage, handling, and transportation. This work evaluates the mechanical properties of Zircaloy-4 samples by means of elevated-temperature three-point bending flexural tests, elevated-temperature microindentation, and elevated-temperature nanoindentation. Experiments...

... To better understand the effect of hydrides on accelerating Zircaloy-4 oxide growth in-reactor, two distinct studies were performed that utilized the Advanced Test Reactor. These and previous studies used three different ways of accelerating the accumulation of hydrogen: the use of samples...

... that the dissolution solvus increases slightly as the hydride volume fraction increases. The apparent hydride volume fraction dependence is judged to result from an increasing self-stress imposed in the material as a result of increasing hydride precipitation. Zircaloy-4 hydride dissolution solvus 780...

... High hydrogen pickup (HPU) and large scatter on HPU at high burnup/high exposure times is observed for Zircaloy-2 (Zr-2), which is used as fuel cladding, fuel channels, water rods, and spacer materials in boiling water reactors (BWRs). However, Zircaloy-4 (Zr-4) does not show this behavior...

... Notched specimens with high or low constraint were tested in situ with a scanning electron microscope using hydrided Zircaloy-4 to develop an understanding of the mechanism for fracture initiation and propagation. High-resolution electron backscatter diffraction was used to identify the texture...

... This article proposes a detailed study of the recrystallization of Zircaloy‐4 under hot forming conditions by means of experimental and numerical tools. Thermomechanical tests and characterization campaigns that have been necessary for this work are described. Then, the different microstructure...

... is given by the author, and an extensive list of references covers the area. The original paper was published by ASTM International in STP 1354, Zirconium in the Nuclear Industry: Twelfth International Symposium , 2000, pp. 15–31. zirconium Zircaloy-2 Zircaloy-4 neutron irradiation irradiation...

... zirconium alloys ZIRLO Zircaloy-4 tin niobium iron corrosion water steam lithium hydroxide in-reactor corrosion microstructure precipitates fuel cladding heat treatment creep growth PWRs RPS2-EB/Nov 2010 Copyright © 2010 by ASTM International 359 www.astm.org 360 ZIRCONIUM PRODUCTION...

... Zircaloy-4 (Zr-4) is a zirconium alloy that has been used extensively within the nuclear industry in both power generation and research reactors. The potential for welded reactor core components to undergo embrittlement during neutron irradiation is an important materials property issue...

... describe hydride nucleation, growth, and coalescence as well as hydride interaction with external stresses. Therefore, an accurate structural characterization was performed on Zircaloy-4 plates and it allowed us to identify a new zirconium hydride phase called ζ. The ζ phase has a trigonal symmetry...

... Sufficient mechanical ductility of high-burnup Zircaloy-4 fuel cladding is important to prevent large-opening ruptures and significant fuel dispersal during postulated in-reactor and spent-fuel processing accidents. The effect of irradiation, oxidation, and hydriding at high fuel burnup may...

... bound on the blister size to have any influence on the rupture pressure. Further, local plastic circumferential strain at each blister can be correlated to relative hydride lens area, as projected onto the cladding surface. mechanical tests hydriding fluence ductility Zircaloy-4 Journal...

... electrochemical impedance spectroscopy (EIS) tests were conducted in the Halden Test Reactor under PWR type conditions. The EIS tests were conducted using five identically fabricated electrode cells, which employed Zircaloy-4 tipped working electrodes. Approximately 30 μn thick oxides were prefilmed...

... by corrosion. Oxide layer and hydrides precipitation could both induce an increase of the dimensional parameters. The scope of the present study is, regardless of irradiation effect, to quantify and to understand the separate effects of hydrogen and oxide layers on the Zircaloy-4 dimensional changes...

... to the final alloy composition. The resulting material, ZIRLO, has proven itself as a successful alloy for PWR fuel cladding and structural applications. zirconium zirconium alloys ZIRLO Zircaloy-4 tin niobium iron corrosion water steam lithium hydroxide in-reactor corrosion microstructure...

... The corrosion and hydriding behavior of M5 and low tin Zircaloy-4 in PWR have been compared at high burn-ups (>70 GWd/t). Whereas low tin Zircaloy-4 reached about 90 µm in oxide thickness and is subjected to significant precipitation of hydride rims at the metal/oxide interface, the oxide...

... Zircaloy fuel cladding suffers progressive degradation of ductility as its neutron exposure and hydrogen uptake increase with burnup. The loss of ductility appears to be the key property governing the cladding integrity in service. We report ductility data of Zircaloy-4 fabricated in stress...

... were conducted for several grid-support/fuel-rod designs using either ZIRLO™, Zircaloy-4, and/or OPTIN™ materials. Rod-to-grid relative displacements and contact forces were measured throughout each test. The wear volumes of each fret mark on the fuel rod specimens were measured via surface...

... The fracture behavior of unirradiated Zircaloy-4 containing either solid hydride blisters or hydrided rims has been examined for the contrasting conditions of equal-biaxial and plane-strain tensile deformation at three temperatures (25°, 300°, and 375°C). Cold-worked and stress-relieved Zircaloy...