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Showing 1 to 20 of 48 for “"Zircaloy"”.
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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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Steam oxidation of FeCrAl coated Zircaloy-2
… calorimetry (DSC). Bare and FeCrAl coated Zircaloy-2 nuclear fuel cladding and FeCrAl coated single crystal Yttria stabilized Zirconia (YSZ) samples were oxidized in an argon/steam environment at 700 C and 1100 C. FeCrAl coated Zircaloy-2 was also heated to 1200 C in an inert environment. …
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An analysis of second phase particles in zircaloy 2
Thesis (M.S.)--Massachusetts Institute of Technology, Dept. of Materials Science and Engineering, 1980.
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Stress corrosion cracking of Zircaloy in the prescence of iodine
Thesis (Sc.D.)--Massachusetts Institute of Technology, Dept. of Materials Science and Engineering, 1981.
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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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Lattice Strain and Texture of Plastically Deformed Zircaloy-2 at 77K
Zircaloy-2 is used extensively in the nuclear industry as a structural material for the reactor core in both light and heavy water reactors. The intergranular strains and texture greatly affect the mechanical properties of the material while in operation. Understanding the plastic deformation of …
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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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Crystallite orientation analysis for zircaloy application of three dimensional representation of textures
Thesis (Nucl.E.)--Massachusetts Institute of Technology, Dept. of Nuclear Engineering, 1981.
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Characterization of Iron-Chromium-Aluminum surface coated Zircaloy-2 under high temperature steam
… (FeCrAl) thin films have been deposited on Zircaloy-2 (Zry2) and yttria-stabilized zirconia (YSZ) by magnetron sputtering, and they have been tested with respect to the oxidation weight gain in high temperature steam. In 700C steam environment, the FeCrAl-coated Zry2 samples have …
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High temperature steam oxidation of titanium-coated Zircaloy-2 and Titanium-Zirconium Alloys
… reaction of high temperature steam and the Zircaloy fuel cladding within the reactor during a Loss-of -Coolant Accident (LOCA). The hydrogen, which collected in the reactor vessel and containment buildings, caused devastating hydrogen gas explosions which lead to the release of radioactive …
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Development of submerged in-situ non-destructive evaluation of hydrogen in Zircaloy fuel cladding
Includes bibliographical references (pages 99-103).
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Response of chromium and chromium-aluminum coatings on zircaloy-2 to high temperature steam
… Our approach is to apply CrAl or Cr coatings on Zircaloy by magnetron sputtering to improve the oxidation resistance of cladding materials. The goal of this research is to characterize the oxidation kinetics and microstructure of CrAl and Cr coatings on Zircaloy-2 in high temperature steam (HTS) …
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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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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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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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