Transmission Electron Microscopy Characterization of Zircaloy-4 Subjected to Ion Irradiation


Book Description

In the operation of a nuclear reactor, the performance of the fuel cladding is critical to ensuring safe and reliable operation of the reactor. The current generation of Light Water Reactors utilizes claddings made from zirconium alloys. The material used for nuclear reactors must be able to withstand temperatures above 3000C while also being exposed to water, high pressures, and radiation. During operation, the zirconium cladding corrodes and picks up hydrogen which can adversely affect its performance. The corrosion mechanisms at work have yet to be fully characterized, especially the influence of irradiation. In order to better understand the mechanisms at work and characterize the behavior of zirconium alloys under reactor conditions, the Mechanistic Understanding of Zirconium Alloy Corrosion (MUZIC) consortium focused on the autoclave corrosion (MUZIC-1) and hydrogen pickup (MUZIC-2) outside of irradiation. The MUZIC-3 effort focuses on corrosion under irradiation. While it would be optimal to test reactor-irradiated samples, the difficulties posed by irradiating, corrosion testing, and examining these samples makes ion irradiation a more appealing manner of irradiation. Using doses and temperatures adjusted for substitution of protons for neutron radiation, this experiment seeks to characterize the effects of irradiation on the base metal, oxide layer, and water, both separately and jointly, on the corrosion of zirconium alloys. In this thesis, the beginning stages of this project, part of MUZIC-3, are presented. This involves verification of the effect of proton irradiation (which is used to represent neutron irradiation) on the base metal and characterization of the irradiated samples. The corrosion testing of this irradiated material will provide a reference for the effect of irradiation induced microstructure changes to the base metal on corrosion. In order to characterize the samples, chemical analyses and observations on crystallinity of secondary phase particles are needed. Along with the analysis of second-phase precipitates, assessment of dislocation loops to observe similarities between different radiation types is also required. Accordingly, samples were irradiated with charged particles (protons and zirconium ions) at the Michigan Ion Beam Laboratory and focused ion beam samples were prepared for transmission electron microscopy examination. The microstructure of the base metal is examined for a range of doses and irradiation temperatures and compared to the microstructure created under neutron irradiation as a preliminary to corrosion testing of irradiated samples. The results are discussed in light of existing literature.




Study of the Initial Stage and Anisotropic Growth of Oxide Layers Formed on Zircaloy-4


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An in situ investigation of the epitaxial oxide layer formed on a thin specimen heated in a transmission electron microscope was carried out. Some dot-like grains about 10 nm in size were formed on the surface of a relatively thin area. The dot-like grains are monoclinic zirconium oxide and have an orientation relationship of (001)m//(01 ̄11)?-Zr, (001)m//(11 ̄01)?-Zr, with the ?-Zr matrix. Some long strip-like grains, probably a new kind of zirconium suboxide, were formed on the surface of a relatively thick area. The strip-like grains have a bcc structure with a lattice parameter a=0.66 nm and have an orientation relationship of (110)bcc//(10110)?-Zr, [1110]bcc//[0001]?-Zr with the ?-Zr matrix. The relationship between the thickness of oxide layers and the grain orientations of the ?-Zr matrix was studied with coarse-grained Zircaloy-4 specimens through autoclave corrosion tests at 500 and 400°C in superheated steam, and at 360°C in both lithiated and deionized water for long time exposure. The results show that the anisotropic growth of oxide layers on the grain surface with different orientations is considerable. However, the relationship between the thickness of oxide layers and the grain orientations of the ?-Zr matrix varies with corrosion temperature and water chemistry. The largest variation of oxide thickness developed during corrosion tests at 500°C. The thickest oxide layers were formed on those grains whose surface orientations were distributed around the planes from (011 ̄0) to (1 ̄21 ̄0). The thicker oxide layers on these grains were further developed into nodular corrosion. When the specimens were corroded at 360°C in lithiated water, the thickest oxide layers formed on those grains, whose surface orientations tilted from 15 to 30° away from the (0001) plane. When the specimens were corroded at 400°C in superheated steam and at 360°C in deionized water, the difference between the thickness of oxide layers on different grain surfaces was less prominent.




Oxidation of Zirconium and Zirconium Alloys


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The oxidation rate was found to be relatively insensitive to various types of surface preparations in the temperature range 400 to 700 deg C. No dependence of reaction rate on oxygen pressure was observed. The cubic rate law also was obeyed by foil specimens at 700 deg C; however, the rate constants were slightly larger than values obtained from parallelepiped samples.







Zirconium in the Nuclear Industry


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Microstructure Evolution in Ion-Irradiated Oxidized Zircaloy-4 Studied with Synchrotron Radiation Microdiffraction and Transmission Electron Microscopy


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The corrosion process (oxidation and hydriding) of zirconium alloy fuel cladding is one of the limiting factors on fuel rod lifetime, particularly for Zircaloy-4. The corrosion rate of this alloy shows indeed a great acceleration at high burnup in light water reactors (LWRs). Understanding the corrosion behavior under irradiation for this alloy is an important technological issue for the safety and efficiency of LWRs. In particular, understanding the effect of irradiation on the metal and oxide layers is a key parameter in the study of corrosion behavior of zirconium alloys. In this study, Zircaloy-4 samples underwent helium and proton ion irradiation up to 0.3 dpa, forming a uniform defect distribution up to 1 ?m deep. Both as-received and precorroded samples were irradiated to compare the effect of metal irradiation to that of oxide layer irradiation. After irradiation, samples were corroded to study the impact of irradiation defects in the metal and in preexisting oxide layers on the formation of new oxide layers. Synchrotron X-ray microdiffraction and microfluorescence were used to follow the evolution of oxide crystallographic phases, texture, and stoichiometry both in the metal and in the oxide. In particular, the tetragonal oxide phase fraction, which has been known to play an important role in corrosion behavior, was mapped in both unirradiated and irradiated metals at the submicron scale and appeared to be significantly affected by irradiation. These observations, complemented with electron microscopy analyses on samples in carefully chosen areas of interest, were combined to fully characterize changes caused by irradiation in metal and oxide phases of both alloys.




Zirconium in the Nuclear Industry


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Zirconium in the Nuclear Industry


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Oxidation and the Testing of Turbine Oils


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This work presents papers from a December 2005 symposium held in Norfolk, Virginia, and sponsored by ASTM Committee D2 on Petroleum Products and Lubricants and its Subcommittees D02.09 on Oxidation and D02.C0 on Turbine Oils. Contributors include equipment manufacturers, end users, lubricant producers, lubricant additive suppliers, test equipment manufacturers, and standard test method developers. They share information on industry trends, evolving technologies, and changing equipment designs and operating conditions, with a focus on how these factors impact oxidation. Some topics covered include turbine oil performance limits, a new form of the rotating pressure vessel oxidation test, and degradation mechanisms leading to sludge and varnish in modern turbine oil formulations. B&w photos are included. There is no subject index. Migdal is affiliated with Chemtura Corporation.