Studies Regarding Corrosion Mechanisms in Zirconium Alloys PDF Download

Are you looking for read ebook online? Search for your book and save it on your Kindle device, PC, phones or tablets. Download Studies Regarding Corrosion Mechanisms in Zirconium Alloys PDF full book. Access full book title Studies Regarding Corrosion Mechanisms in Zirconium Alloys by M. Preuss. Download full books in PDF and EPUB format.

Studies Regarding Corrosion Mechanisms in Zirconium Alloys

Studies Regarding Corrosion Mechanisms in Zirconium Alloys PDF Author: M. Preuss
Publisher:
ISBN:
Category : Corrosion mechanisms
Languages : en
Pages : 23

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.

Studies Regarding Corrosion Mechanisms in Zirconium Alloys

Studies Regarding Corrosion Mechanisms in Zirconium Alloys PDF Author: M. Preuss
Publisher:
ISBN:
Category : Corrosion mechanisms
Languages : en
Pages : 23

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.

Zirconium in the Nuclear Industry

Zirconium in the Nuclear Industry PDF Author: George P. Sabol
Publisher: ASTM International
ISBN: 0803124066
Category : Nuclear fuel claddings
Languages : en
Pages : 907

Book Description


Understanding of Corrosion Mechanisms of Zirconium Alloys After Irradiation

Understanding of Corrosion Mechanisms of Zirconium Alloys After Irradiation PDF Author: Marc Tupin
Publisher:
ISBN:
Category : Corrosion
Languages : en
Pages : 41

Book Description
The irradiation damage in the fuel cladding material is mainly caused by the neutron flux resulting from the fission reactions occurring in the fuel. From an experimental point of view, the neutrons have the disadvantage to activate materials by neutron capture rendering them difficult to handle. To avoid these constraints inherent in the handling of radioactive material, the radiation effects on the corrosion resistance of zirconium alloys can be studied by irradiating the materials with ions. A new experimental approach using ion irradiation was performed in the Microscopy and Irradiation Damage Studies Laboratory of the CEA in Saclay, with the aim to study more specifically the influence of the irradiation damages in the oxide on the corrosion rate of the zirconium alloys. This study was, moreover, focused on a particular distribution of defects in the oxide layer, basically, localised close to the metal/oxide interface. From the results of the irradiation of the metal/oxide interface, it was clearly shown that, whatever the incident ion, the irradiation of the internal interface results in a significant increase of the oxygen diffusion flux ratios between the most irradiated Zircaloy-4 and the unirradiated one, whereas that of the oxide formed on M5TM induces a big decrease of the oxygen diffusion flux in the film. These effects are less marked with helium ions compared to protons (M5TM is a trademark of AREVA NP registered in the United States and in other countries). Finally, the oxide irradiation impact on the oxygen diffusion through the layer could explain the corrosion acceleration factor observed on Zy4 during the first cycles of irradiation, but cannot alone explain observed corrosion accelerations under high burn-up conditions. The discussion on the oxide irradiation effects puts forward the probable role of the residual charge left by ion implantation.

Studies of the Hydrogen Damage Mechanism in the Corrosion of Zirconium

Studies of the Hydrogen Damage Mechanism in the Corrosion of Zirconium PDF Author: R. D. Misch
Publisher:
ISBN:
Category : Corrosion and anti-corrosives
Languages : en
Pages : 46

Book Description
In complex alloys, cathodes may arise in a variety of ways and the corrosion resistance of the intermetallics may be of importance.

Examinations of the Corrosion Mechanism of Zirconium Alloys

Examinations of the Corrosion Mechanism of Zirconium Alloys PDF Author: HJ. Beie
Publisher:
ISBN:
Category : Barrier layer
Languages : en
Pages : 29

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.

Corrosion of Zirconium Alloys in 900° Steam

Corrosion of Zirconium Alloys in 900° Steam PDF Author: J. Paul Pemsler
Publisher:
ISBN:
Category : Steam
Languages : en
Pages : 26

Book Description


Oxidation of Zirconium and Zirconium Alloys

Oxidation of Zirconium and Zirconium Alloys PDF Author:
Publisher:
ISBN:
Category : Oxidation
Languages : en
Pages : 48

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.

Corrosion Mechanism of Zirconium and Its Alloys

Corrosion Mechanism of Zirconium and Its Alloys PDF Author: David Leslie Douglass
Publisher:
ISBN:
Category : Zirconium
Languages : en
Pages : 32

Book Description


Zirconium in the Nuclear Industry

Zirconium in the Nuclear Industry PDF Author: Gerry D. Moan
Publisher: ASTM International
ISBN: 0803128959
Category : Nuclear fuel claddings
Languages : en
Pages : 891

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.

On the Initial Corrosion Mechanism of Zirconium Alloy

On the Initial Corrosion Mechanism of Zirconium Alloy PDF Author: U. Döbler
Publisher:
ISBN:
Category : Corrosion
Languages : en
Pages : 19

Book Description
The initial stages of zirconium oxide formation on Zircaloy after water (H2O) and oxygen (O2) exposures have been investigated in situ using photoelectron spectroscopy and X-ray-absorption spectroscopy. The reactivity of the zirconium alloy with O2 at room temperature is about 1000 times higher than for H2O. Up to 100 L (1 L = 1 Langmuir unit = 1 • 10-6 mbar • s) H2O exposure, the reactivity of the zirconium alloy at 450°C is comparable to the room temperature reaction. At higher H2O exposure, a sharp increase in the reaction rate for the high-temperature oxidation is observed. From the energy position of the Zr 3d photo emission line and their oxygen-induced chemical shifts, one can directly follow the formation of the oxide films. Two different substoichiometric oxides were found during reaction with water. Suboxide (1) is located at the zirconium/zirconium-oxide interface. Subsequently, a Suboxide (2) is concluded from the chemical shift of the zirconium photoelectrons. After an oxide thickness of 2 nm, the stoichiometric ZrO2 phase is not yet developed.