Download or read book Irradiation Effects on the Mechanical Properties of Zirconium and Dilute Zirconium Alloys written by and published by . This book was released on 1976 with total page pages. Available in PDF, EPUB and Kindle. Book excerpt: The effects of fast flux (E greater than or equal to 1 MeV) neutrons on zirconium and dilute zirconium alloys are discussed. The effects on the elastic constants, strength and ductility, creep, fatigue and fracture, and irradiation growth are reviewed. (FS).
Download or read book Mechanical Properties of Irradiated Zirconium Zircaloy and Aluminum written by Richard E. Schreiber and published by . This book was released on 1961 with total page 112 pages. Available in PDF, EPUB and Kindle. Book excerpt:
Download or read book The Effect of Irradiation on the Mechanical Properties of Arc melted Bureau of Mines Zirconium with Various Degrees of Cold Work written by R. S. Kemper and published by . This book was released on 1955 with total page 68 pages. Available in PDF, EPUB and Kindle. Book excerpt:
Download or read book The Effect of Neutron Irradiation on the Mechanical Properties of Hydrided Zirconium Alloys written by Atomic Energy of Canada Limited and published by . This book was released on 1964 with total page pages. Available in PDF, EPUB and Kindle. Book excerpt:
Download or read book Effect of Irradiation at 588 K on Mechanical Properties and Deformation Behavior of Zirconium Alloy Strip written by J. Baicry and published by . This book was released on 1987 with total page 19 pages. Available in PDF, EPUB and Kindle. Book excerpt: Irradiation tests conducted for the Advanced Fuel Assembly (AFA) grid development allowed the measurement of the mechanical characteristics, growth, and relaxation of the following materials: stress-relieved, ?-annealed, and ?-quenched Zircaloy-4, and ?-annealed Zr-3Sn-1Mo. The mechanical characteristics of the zirconium alloys approached a limit value as early as 5 x 1024 n m-2, except for the ?-quenched Zircaloy-4 for which the limit value was reached later at about 5 x 1025 n m-2. Uniform elongation at 588 K showed a minimum at about 5 x 1024 n m-2; beyond this fluence, stress-relieved Zircaloy-4 presented the most marked tendency to increase. Longitudinal direction growth of annealed Zircaloy-4 and longitudinal and transverse direction growth of stress-relieved Zircaloy-4 may be expressed by the formula ? = A(?t)n where n = 0.67 (stress-relieved Zircaloy-4) and n = 0.4 (annealed Zircaloy-4). It is doubtful that the growth is associated with a density change. The stress-relaxation of the zirconium base materials is nearly complete for 4 x 1025 n m-2, with the exception of Zr-1Nb which has an intermediate behavior between those of Zircaloy-4 and Inconel 718.
Download or read book NUREG CR written by U.S. Nuclear Regulatory Commission and published by . This book was released on 1979 with total page 160 pages. Available in PDF, EPUB and Kindle. Book excerpt:
Download or read book In PWR Irradiation Performance of Dilute Tin Zirconium Advanced Alloys written by GP. Smith and published by . This book was released on 2002 with total page 17 pages. Available in PDF, EPUB and Kindle. Book excerpt: Zirconium alloys containing about 0.5% tin, which are classified as dilute tin alloys, possess excellent uniform waterside corrosion resistance necessary for the PWR fuel applications. Mechanical and irradiation growth properties of the dilute alloys can be adjusted for specific component application by controlling the additions of other alloying elements such as iron, chromium, niobium, and oxygen. Cladding alloys with such additions have been successfully irradiated to burnups up to 69 GWd/MTU, showing significant improvements in corrosion resistance and irradiation growth characteristics compared to low-tin Zircaloy-4, one of the current standard materials. The in-PWR creep resistance of such dilute alloys is comparable to that of low-tin Zircaloy-4. Another dilute alloy with predominantly iron-containing second-phase particles that are unstable under neutron irradiation (in a cold-worked microstructure, cold work introduced prior to irradiation) appears to be most suitable for the grid strip application. Cold-worked I-spring of this alloy in a transverse stamped grid provides excellent fuel rod support by inward motion of the spring within the grid cell due to irradiation growth. The hydrogen pickup fraction of several zirconium alloys, including Zircaloy-4 and dilute alloys, exhibits a well-behaved correlation with oxide thickness under non-heat flux conditions. A similar correlation is expected under heat flux conditions. Under heat flux conditions, the hydrogen pickup fraction for Zircaloy-4 approaches a constant value of about 15% for oxide thicknesses greater than 50 ?m. For the non heat-flux conditions, the pickup fraction is less than 5% for oxide thickness values greater than 50 ?m. Possible reasons for influence of oxide thickness and heat flux on the hydrogen pickup fraction are the porosity traps in thick oxide layers and atomic vibrations of oxide lattice under heat flux conditions. The in-PWR performance characteristics of the dilute alloys such as corrosion resistance, ductility, and dimensional stability can be controlled by optimization of the composition and fabrication process. These parameters influence the composition of the second-phase particles (SPP) in the alloy microstructure, which determines the radiation stability of the SPP. Irradiation stabilityof SPP has strong impact on the in-PWR performance characteristics of zirconium alloys.
Download or read book Impact of Irradiation Damage Recovery During Transportation on the Subsequent Room Temperature Tensile Behavior of Irradiated Zirconium Alloys written by B. Bourdiliau and published by . This book was released on 2010 with total page 25 pages. Available in PDF, EPUB and Kindle. Book excerpt: Zirconium alloys are commonly used in pressurized water reactor as fuel rod cladding tubes. After irradiation and cooling in pool, the spent nuclear fuel assemblies are either transported for wet storage to a devoted site or loaded in suitable casks for dry storage on a nuclear power plant site. During dry transportation or at the beginning of dry storage, at temperatures around 400°C, the cladding experiences a creep deformation under the hoop stress induced by the internal pressure of the fuel rod. During creep, a recovery of the radiation damage can occur that can affect the subsequent mechanical properties. The mechanical behavior of the cladding has been investigated in laboratory on two neutron irradiated cladding materials: fully recrystallized Zr-1 % Nb and stress-relieved Zircaloy-4. Creep tests under internal pressure were conducted at 400 and 420°C. After depressurization and cooling, ring tensile tests were carried out at room temperature. In addition, transmission electron microscopy observations have been performed after testing. The post-creep mechanical response exhibited a decrease of the strength compared to the as-irradiated material. This decrease is associated with a significant recovery of the ductility, which becomes close to the ductility of the unirradiated material. The transmission electron microscopy examinations, conducted on recrystallized Zr-1 % Nb ring samples, revealed that the radiation defects have been annealed. It was also observed that as for the unirradiated material, the deformation occurred homogeneously throughout the grains. No dislocation channeling was observed contrary to the as-irradiated material. These observations explain the recovery of the strength and of the ductility after post-irradiation creep that may also occur during dry transportation or at the beginning of dry storage.
Download or read book Energy Research Abstracts written by and published by . This book was released on 1995 with total page 782 pages. Available in PDF, EPUB and Kindle. Book excerpt:
Download or read book Zirconium in the Nuclear Industry written by ASTM Committee B-10 on Reactive and Refractory Metals and Alloys and published by ASTM International. This book was released on 1977 with total page 694 pages. Available in PDF, EPUB and Kindle. Book excerpt:
Download or read book Nuclear Science Abstracts written by and published by . This book was released on 1976 with total page 612 pages. Available in PDF, EPUB and Kindle. Book excerpt:
Download or read book Scientific and Technical Aerospace Reports written by and published by . This book was released on 1970 with total page 348 pages. Available in PDF, EPUB and Kindle. Book excerpt:
Download or read book Applications Related Phenomena in Zirconium and Its Alloys written by Committee B-10 Staff and published by ASTM International. This book was released on 1969 with total page 393 pages. Available in PDF, EPUB and Kindle. Book excerpt:
Download or read book The Effect of Irradiation on the Impact Properties of Some Zirconium Alloys written by D.S. Wood and published by . This book was released on 1965 with total page 11 pages. Available in PDF, EPUB and Kindle. Book excerpt:
Download or read book Mechanistic Considerations Used in the Development of the PROFIT PCI Failure Model written by P. J. Pankaskie and published by . This book was released on 1980 with total page 92 pages. Available in PDF, EPUB and Kindle. Book excerpt:
Download or read book Quantifying Irradiation Defects in Zirconium Alloys written by Levente Balogh and published by . This book was released on 2018 with total page 34 pages. Available in PDF, EPUB and Kindle. Book excerpt: Irradiation-induced dislocations significantly affect the mechanical properties of zirconium alloys, altering slip and influencing creep and growth. Thus, the quantitative characterization of irradiation defects as a function of fluence, cold work, and/or thermal treatments is important for models that attempt to predict their impact on properties. Whole-pattern diffraction line-profile analysis (DLPA) is a well-established modern tool for microstructure characterization based on first-principle physical models for dislocation density measurements in plastically deformed materials. However, applying these DLPA methods directly to irradiated materials yields higher than expected dislocation density values compared with historical transmission electron microscopy (TEM) measurements and past line-broadening analysis studies calibrated to TEM observations. In an effort to understand these differences, a new microstructural model was developed for DLPA to specifically address dislocation structures consisting of elliptical a- and c-component loops. To compare the refined DLPA method with TEM measurements, high-resolution neutron diffraction patterns on nonirradiated and irradiated Zr-2.5Nb samples were collected with the Neutron Powder Diffractometer instrument at the Los Alamos Neutron Science Center and were evaluated. High-resolution TEM measurements were performed at the Reactor Materials Testing Laboratory, Queen's University, for comparison with the DLPA results. The capabilities and inherent uncertainties of both the refined DLPA and TEM methods are compared and discussed in detail. We show that the differences between the density values provided by DLPA and TEM are inherent to the methods and can be reconciled with the interpretation of the data.
Download or read book The Effect of Corrosion and Fast Neutron Irradiation on the Mechanical Properties of the Zirconium 8 6 Wt Aluminum Reference Material written by Eugene V. T. Murphy and published by Pinawa, Man. : Materials and Component Development Branch, Whiteshell Nuclear Research Establishment. This book was released on 1978 with total page 19 pages. Available in PDF, EPUB and Kindle. Book excerpt:
