Examinations of the Corrosion Mechanism of Zirconium Alloys PDF Download
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Author: HJ. Beie
Publisher:
ISBN:
Category : Barrier layer
Languages : en
Pages : 29
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Book Description
Several mechanism-related aspects of the corrosion of zirconium alloys have been investigated using different examination techniques. The microstructure of different types of oxide layers was analyzed by transmission electron microscopy (TEM). Uniform oxide mainly consists of m-ZrO2 and a smaller fraction of t-ZrO2 with columnar grains and some amount of equiaxed crystallites. Nodular oxides show a high open porosity and the grain shape tends to the equiaxed type. A fine network of pores along grain boundaries was found in oxides grown in water containing lithium. An enrichment of lithium within such oxides could be found by glow discharge optical spectroscopy (GD-OES) depth profiling. In all oxides, a compact, void-free oxide layer was observed at the metal/oxide interface. Compressive stresses within the oxide layer measured by an X-ray diffraction technique were significantly higher compared to previously published values. Electrical potential measurements on oxide scales showed the influence of the intermetallic precipitates on the potential drop across the oxide. In long-time corrosion tests of Zircaloy with varying temperatures, memory effects caused by the cyclic formation of barrier layers could be observed. It was concluded that the corrosion mechanism of zirconium-based alloys is a barrier-layer controlled process. The protective properties of this barrier layer determine the overall corrosion resistance of zirconium alloys.
Author: HJ. Beie
Publisher:
ISBN:
Category : Barrier layer
Languages : en
Pages : 29
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Book Description
Several mechanism-related aspects of the corrosion of zirconium alloys have been investigated using different examination techniques. The microstructure of different types of oxide layers was analyzed by transmission electron microscopy (TEM). Uniform oxide mainly consists of m-ZrO2 and a smaller fraction of t-ZrO2 with columnar grains and some amount of equiaxed crystallites. Nodular oxides show a high open porosity and the grain shape tends to the equiaxed type. A fine network of pores along grain boundaries was found in oxides grown in water containing lithium. An enrichment of lithium within such oxides could be found by glow discharge optical spectroscopy (GD-OES) depth profiling. In all oxides, a compact, void-free oxide layer was observed at the metal/oxide interface. Compressive stresses within the oxide layer measured by an X-ray diffraction technique were significantly higher compared to previously published values. Electrical potential measurements on oxide scales showed the influence of the intermetallic precipitates on the potential drop across the oxide. In long-time corrosion tests of Zircaloy with varying temperatures, memory effects caused by the cyclic formation of barrier layers could be observed. It was concluded that the corrosion mechanism of zirconium-based alloys is a barrier-layer controlled process. The protective properties of this barrier layer determine the overall corrosion resistance of zirconium alloys.
Author: Leo F. P. Van Swam
Publisher: ASTM International
ISBN: 0803111991
Category : Nuclear fuel claddings
Languages : en
Pages : 781
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Book Description
Author: George P. Sabol
Publisher: ASTM International
ISBN: 0803124066
Category : Nuclear fuel claddings
Languages : en
Pages : 907
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Book Description
Author: F. Garzarolli
Publisher:
ISBN:
Category : Characterization of oxide layers
Languages : en
Pages : 21
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Book Description
In order to get a better understanding of the mechanisms governing corrosion of Zr-based alloys, several examinations have been performed on a variety of samples with uniform and nodular corrosion and different oxide layer thicknesses.
Author: George P. Sabol
Publisher: ASTM International
ISBN: 0803124996
Category : Microstructure
Languages : en
Pages : 953
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Book Description
Author: M. Preuss
Publisher:
ISBN:
Category : Corrosion mechanisms
Languages : en
Pages : 23
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Book Description
Understanding the key corrosion mechanisms in a light water reactor primary water environment is critical to developing and exploiting improved zirconium alloy fuel cladding. In this paper, we report recent research highlights from a new collaborative research programme involving 3 U.K. universities and 5 partners from the nuclear industry. A major part of our strategy is to use the most advanced analytical tools to characterise the oxide and metal/oxide interface microstructure, residual stresses, as well as the transport properties of the oxide. These techniques include three-dimensional atom probe (3DAP), advanced transmission electron microscopy (TEM), synchrotron X-ray diffraction, Raman spectroscopy, and in situ electro-impedance spectroscopy. Synchrotron X-ray studies have enabled the characterisation of stresses, tetragonal phase fraction, and texture in the oxide as well as the stresses in the metal substrate. It was found that in the thick oxide (here, Optimized-ZIRLO, a trademark of the Westinghouse Electric Company, tested at 415°C in steam) a significant stress profile can be observed, which cannot be explained by metal substrate creep alone but that local delamination of the oxide layers due to crack formation must also play an important role. It was also found that the oxide stresses in the monoclinic and tetragonal phases grown on Zircaloy-4 (autoclave testing at 360°C) first relax during the pre-transition stage. Just before transition, the compressive stress in the monoclinic phase suddenly rises, which is interpreted as indirect evidence of significant tetragonal to monoclinic phase transformation taking place at this stage. TEM studies of pre- and post-transition oxides grown on ZIRLO, a trademark of the Westinghouse Electric Company, have used Fresnel contrast imaging to identify nano-sized pores along the columnar grain boundaries that form a network interconnected once the material goes through transition. The development of porosity during transition was further confirmed by in situ electrochemical impedance spectroscopy (EIS) studies. 3DAP analysis was used to identify a ZrO sub-oxide layer at the metal/oxide interface and to establish its three-dimensional morphology. It was possible to demonstrate that this sub-oxide structure develops with time and changes dramatically around transition. This observation was further confirmed by in situ EIS studies, which also suggest thinning of the sub-oxide/barrier layer around transition. Finally, 3DAP analysis was used to characterise segregation of alloying elements near the metal/oxide interface and to establish that the corroding metal near the interface (in this case ZIRLO) after 100 days at 360°C displays a substantially different chemistry and microstructure compared to the base alloy with Fe segregating to the Zr/ZrO interface.
Author: J. Paul Pemsler
Publisher:
ISBN:
Category : Steam
Languages : en
Pages : 26
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Book Description
Author: Craig M. Eucken
Publisher: ASTM International
ISBN: 080311463X
Category : Nuclear fuel claddings
Languages : en
Pages : 794
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Book Description
The proceedings of the Ninth International Symposium on [title], held in Kobe, Japan, November 1990, address current trends in the development, performance, and fabrication of zirconium alloys for nuclear power reactors. the bulk of the most recent work on zirconium alloy behavior has concerned corr
Author:
Publisher:
ISBN:
Category : Oxidation
Languages : en
Pages : 48
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Book Description
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.
Author: Gerry D. Moan
Publisher: ASTM International
ISBN: 0803128959
Category : Nuclear fuel claddings
Languages : en
Pages : 891
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Book Description
Annotation The 41 papers of this proceedings volume were first presented at the 13th symposium on Zirconium in the Nuclear Industry held in Annecy, France in June of 2001. Many of the papers are devoted to material related issues, corrosion and hydriding behavior, in-reactor studies, and the behavior and properties of Zr alloys used in storing spent fuel. Some papers report on studies of second phase particles, irradiation creep and growth, and material performance during loss of coolant and reactivity initiated accidents. Annotation copyrighted by Book News, Inc., Portland, OR.
