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Showing 1 to 16 of 16 for “"Zircaloy-4"”.
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Stress relaxation of Zircaloy-4.
Thesis. 1976. B.S.--Massachusetts Institute of Technology. Dept. of Materials Science and Engineering.
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Hydrogen entry in Zircaloy-4 fuel cladding : an electrochemical study
… of hydrogen liberated by waterside corrosion of Zircaloy-4 fuel cladding in Pressurized Water Reactors (PWRs): are the adsorbed protons incorporated into the oxide and eventually the metal, or are they evolved into molecular hydrogen and released into the coolant? Water chemistry modeling was …
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Formation of crystallographic textures during manufacturing of tubes of Zircaloy 4.
Thesis. 1977. M.S.--Massachusetts Institute of Technology. Dept. of Materials Science and Engineering.
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Developing an elevated temperature Pourbaix diagram for zirconium and multi-element Pourbaix diagram of Zircaloy-4
Zirconium and Zircaloy-4 are used in applications that require corrosion resistance under extreme conditions. For example, the fuel bundles in coolant systems of CANDU™ reactors operate between 533 to 583K (260 to 310 °C). However, thermodynamic and hydrolysis properties of aqueous zirconium …
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Capturing irradiation-enhanced corrosion of zircaloy-4 with a charge-based diffusion/drift phase field model
Zircaloy-4 has been used in pressurized water reactors (PWRs) for decades, and enhanced corrosion rates in reactors compared to out-of-pile have long been observed. However, the exact mechanism explaining the early departure from autoclave kinetics after 3-5 microns of oxide have formed has proved …
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Investigations on physical vapor deposition chromium coatings on Zircaloy-4 cladding using magnetron sputtering for accident tolerant fuel applications
Submission original under an indefinite embargo labeled 'Open Access'. The submission was exported from vireo on 2025-03-28 without embargo terms
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Influence of zirconium hydride on the biaxial creep behavior of Zircaloy-4 cladding for interim dry storage of spent nuclear fuel
Submission published under a 24 month embargo labeled 'U of I Access', the embargo will last until 2019-12-01
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Investigation of the mechanical behavior of rim/blister structure hydride phases in Zircaloy-4 by utilizing high energy synchrotron X-ray diffraction
Zircaloy-4 is one of the most common fuel cladding materials used in light water reactors due to excellent mechanical properties, corrosion resistance and low thermal neutron absorption cross-section. However, in service absorption of hydrogen by Zircaloy-4 (Zry-4) cladding is potentially …
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Investigation effect of external stress on hydrogen solvus in a zircaloy-4 fuel cladding alloy: application of in situ diffraction technique and mesoscale phase-field simulation
Zircaloy-4 is widely used as a nuclear fuel cladding material in light water reactors (LWRs). However, the formation of the zirconium hydride phase in the cladding significantly degrades its mechanical properties by reducing the local ductility and initiating crack propagation. Therefore, one of …
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Silicon carbide oxidation in high temperature steam
… Nuclear Regulatory Commission is the exothermic zircaloy-steam oxidation reaction which can potentially occur during a loss of coolant accident (LOCA), and may become autocatalytic beyond 1,200 0C, thus generating a large amount of hydrogen. The concern for the zircaloy oxidation reaction has …
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Cracking of zirconium alloys at constant load with and without applied potentials
Cladding materials, namely Zircaloy-2 (Zr-2) and Zircaloy-4 (Zr-4), were tested for evaluation of their resistance to stress corrosion cracking (SCC), delayed hydride cracking (DHC), and localized corrosion (pitting and crevice) in neutral and acidic environments at ambient temperature and 90°C. …
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Evaluation of cracking and localized corrosion behavior of cladding materials
… (pitting and crevice) susceptibility of zircaloy-2 and zircaloy-4 in neutral and acidic solutions at 30, 60, 90°C. Constant-load and slow-strain-rate (SSR) testing methods were used to evaluate the SCC and HE behavior of both alloys by use of smooth and notched tensile specimens. …
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Investigation of Separate Effects of Surface Condition on Subcooled Flow Boiling Heat Transfer
… boiling heat transfer. To that end, we prepared Zircaloy-4 heaters that mimic the commercial PWRs fuel cladding, and conducted subcooled flow boiling experiments at 1 bar, 10 K subcooling, and 1000 kg/m2 s mass flux, using high-resolution high-speed video (HSV) and infrared (IR) diagnostics. A …
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Reactor physics considerations for implementing silicon carbide cladding into a PWR environment
… dimensions remained similar to the current Zircaloy 4 cladding. Three approaches were followed in creating the PWR core designs: 1) constant core power density (or total reactor power) and cycle length, but fewer fresh assemblies loaded, 2) constant cycle length, but increased core power …
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Modelización microscópica del daño por irradiación en metales hcp (a-Zirconio) : efecto de la anisotropía en la difusión y evolución de defectos
… We have focused on zirconium alloys claddings (Zircaloy-4 and Zr-2.5%Nb). The first step has been to simulate the neutron spectra. Spectra for current pressure water reactors (PWR) and high burn-up advanced reactors have been obtained. Neutron spectra in several burnt steps have also been …