Evolution of microstructure in zirconium alloys during irradiation PDF Download
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Author: M. Griffiths
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Category :
Languages : en
Pages : 0
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Author: M. Griffiths
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ISBN:
Category :
Languages : en
Pages : 0
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Author: M. Griffiths
Publisher:
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Category : Congress
Languages : en
Pages : 23
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X-ray diffraction (XRD) and transmission electron microscopy (TEM) have been used to characterize microstructural and microchemical changes produced by neutron irradiation in zirconium and zirconium alloys. Zircaloy-2, Zircaloy-4, and Zr-2.5Nb alloys with differing metallurgical states have been analyzed after irradiation for neutron fluences up to 25 x 1025 n.m-2 (E > 1 MeV) for a range of temperatures between 330 and 580 K.
Author: M. Griffiths
Publisher:
ISBN: 9780803170186
Category : Cavities
Languages : en
Pages : 10
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The performance of zirconium alloys in BWR, PWR, and PHWR nuclear reactors is dependent on the microstructure. Accordingly, the characterization of the microstructure is an integral part of any study conducted to develop models for in-reactor performance. Although the as-fabricated microstructure (texture, grain size, dislocation density, and phase or precipitate distribution) determines the basic physical properties of a given component, there are changes that occur during irradiation that can have a significant effect on these properties. Microstructures that illustrate specific features of the radiation damage that forms in Zr alloys will be illustrated and discussed in terms of the dose, dose rate, and impurity factors that are applicable. The original paper was published by ASTM International in the Journal of ASTM International, December 2007.
Author: M. Griffiths
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Category :
Languages : en
Pages : 0
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Author: M. Griffiths
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Category :
Languages : en
Pages : 0
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Author: PCK Chow
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Category : Corrosion
Languages : en
Pages : 34
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X-ray diffraction (XRD) and analytical electron microscopy (AEM) have been used to characterise microstructural and microchemical changes produced by neutron irradiation of Zr-2.5Nb, Zircaloy-2 and Zircaloy-4 nuclear reactor core components.
Author: GP. Kobylyansky
Publisher:
ISBN:
Category : Dislocation
Languages : en
Pages : 17
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Zirconium alloy components subjected to long-term operation and high doses in thermal reactors need to be highly irradiation resistant to provide integrity of components, primarily, their geometrical sizes. Transmission and scanning electron microscopy, energy dispersive X-ray microanalysis used to investigate thin foils and extraction replicas of irradiated zirconium, Zr-1Nb (E110) and Zr-1.2Sn-1Nb-0.4Fe (E635) alloys allowed us to analyze the evolution of their microstructure under neutron irradiation. The experimental irradiations that were conducted at 350°C to 1027 n/m2 (E >= 0.1 MeV) show that the most irradiation resistant alloy proved to be a multicomponent E635 alloy. It is not essentially subject to growth. Dislocation structure and phase composition were studied as interrelated to different stages of irradiation induced growth. The accelerated growth correlates with a high density of basal -- plane c-component dislocations.
Author: George P. Sabol
Publisher: ASTM International
ISBN: 0803124066
Category : Nuclear fuel claddings
Languages : en
Pages : 907
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Author: Christopher Ulmer
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ISBN:
Category :
Languages : en
Pages :
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ZrC is a candidate material for use in Generation IV high-temperature, gas-cooled reactor TRISO coated fuel particles, so it is important to understand its behavior under irradiation. The microstructural evolution of ZrC$_x$ under irradiation was studied in situ using the Intermediate Voltage Electron Microscope (IVEM) at Argonne National Laboratory. Experiments were performed in which the sample stoichiometry and irradiation temperature were systematically varied. In situ experiments made it possible to continuously follow the microstructure during irradiation using diffraction contrast imaging. Images and diffraction patterns were methodically recorded at chosen dose points. Experiments centered on the irradiation of ZrC$_{0.8}$ and ZrC$_{0.9}$ with 1 MeV Kr ions at temperatures ranging from 20 - 1073 K up to 10 dpa.Initial damage developed as 2 - 4 nm diameter black-dot defects after a threshold dose of approximately 0.1 - 0.5 dpa. As the irradiation temperature increased, the threshold dose for visible defect formation decreased. The density and size of defects increased with additional dose and the density of defects ranged on the order of $10^{22}$ - $10^{23}$ m$^{-3}$ for all experiments. The defect diameter also increased with irradiation temperature, with average defect diameters at 3 dpa ranging from approximately 4 nm at 673 K to 8 nm at 1073 K. No long-range migration of the visible defects or dynamic defect creation and elimination were observed during irradiation, but agglomeration of small defects into loops occurred at 1073 K and resulted in an overall coarsening of the microstructure. The irradiated microstructure was found to not be strongly dependent on the stoichiometry as results for the two stoichiometries studied were nearly identical. No irradiation induced amorphization was observed, even after 5 dpa at 20 K and 10 dpa at 50 K. At the higher temperature (873 K and above), the irradiated microstructure varied with sample thickness and showed a defect-denuded zone in the thin area near the edge.A one-dimensional cluster dynamics rate theory model that only considered the creation and mobility of point defects and their agglomeration into defect clusters was solved and compared with the experimental results. General trends from the simulation results matched the experimental observations: a threshold dose was predicted by the calculation, loop diameter was predicted to increases with dose and temperature, and loop density increased with dose and decreased with temperature, as observed. The spatial distribution showed lower loop size and density near the surface. Additional work is needed to match the experimental results quantitatively for both loop size and density, and the results were found to be sensitive to the chosen temperature.
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.
