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Author: Christopher Ulmer
Publisher:
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: Christopher Ulmer
Publisher:
ISBN:
Category :
Languages : en
Pages :
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Book Description
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: J. Gan
Publisher:
ISBN:
Category : Dislocation
Languages : en
Pages : 7
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The gas-cooled fast reactor (GFR) is one of six concepts for the Generation-IV nuclear energy system. The fuel for the GFR requires both a high heavy metal loading and the ability to withstand temperatures up to 1600°C during a loss of coolant accident. ZrC is among the few potential refractory ceramic materials with necessary properties to be considered as matrix materials for a dispersed carbide fuel. The radiation response of ZrC to high dose and temperature is a critical research need. This work investigated the microstructure of ZrC irradiated with 1 MeV Kr ions to doses of 10 and 30 dpa at 27°C and 10 and 70 dpa at 800°C with a damage rate approximately 3.0 × 10−3 dpa/s. No radiation-induced amorphization was found. A lattice expansion of approximately 7 % was observed for ZrC irradiated to 70 dpa at 800°C.
Author: Raul Fernando Florez Meza Florez
Publisher:
ISBN:
Category :
Languages : en
Pages : 179
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"A comprehensive understanding of the microstructural evolution of Zirconium Carbide (ZrC[sub x]) ceramics under irradiation conditions is required for their successful implementation in advanced Gen-IV gas-cooled nuclear reactors. The research presented in this dissertation focusses on elucidating the ion and electron irradiation response of ZrC[sub x] ceramics. In the first part of the research, the microstructural evolution was characterized for ZrC[sub x] ceramics irradiated with 10 MeV Au3+ ions up to doses of 30 displacement per atom (dpa) at 800 °C. Coarsening of the defective microstructure, as a function of dose, was revealed by transmission electron microscopy analysis. The lack of change in the irradiated microstructure, at doses above 5 dpa, indicated that a balance between irradiation damage accumulation and dynamic annealing of defects was reached. It was also found that concurrent oxidation occurred during the ion irradiation. The effects of irradiation on the morphology and microstructure of the initial oxide formed on the surface of ZrC[sub x] were investigated. The concomitant reduction in size and surface coverage of the oxide nodules at high doses, indicated that oxide dissolution was the predominant mechanism under irradiation conditions. In the second part of the research, Zirconium Carbide (ZrC[sub x]) was irradiated with 10 MeV Au3+ ions to a dose of 10 dpa and subsequently with 300 keV electrons in a transmission electron microscope (TEM). It was found that high-energy electron irradiation of pre-damaged ZrC[sub x] foils induce atomic mixing via radiation enhanced diffusion (RED), producing surface oxidation of the TEM foil"--Abstract, page iv.
Author:
Publisher:
ISBN:
Category :
Languages : en
Pages :
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Zirconium carbide (ZrC) is being considered for utilization in high-temperature gas-cooled reactor fuels in deep-burn TRISO fuel. Zirconium carbide possesses a cubic B1-type crystal structure with a high melting point, exceptional hardness, and good thermal and electrical conductivities. The use of ZrC as part of the TRISO fuel requires a thorough understanding of its irradiation response. However, the radiation effects on ZrC are still poorly understood. The majority of the existing research is focused on the radiation damage phenomena at higher temperatures (>450°C) where many fundamental aspects of defect production and kinetics cannot be easily distinguished. Little is known about basic defect formation, clustering, and evolution of ZrC under irradiation, although some atomistic simulation and phenomenological studies have been performed. Such detailed information is needed to construct a model describing the microstructural evolution in fast-neutron irradiated materials that will be of great technological importance for the development of ZrC- based fuel. The goal of the proposed project is to gain fundamental understanding of the radiation-induced defect formation in zirconium carbide and irradiation response (ZrC) by using a combination of state-of-the-art experimental methods and atomistic modeling. This project will combine (1) in situ ion irradiation at a specialized facility at a national laboratory, (2) controlled temperature proton irradiation on bulk samples, and (3) atomistic modeling to gain a fundamental understanding of defect formation in ZrC. The proposed project will cover the irradiation temperatures from cryogenic temperature to as high as 800°C, and dose ranges from 0.1 to 100 dpa. The examination of this wide range of temperatures and doses allows us to obtain an experimental data set that can be effectively used to exercise and benchmark the computer calculations of defect properties. Combining the examination of radiation-induced microstructures mapped spatially and temporally, microstructural evolution during post-irradiation annealing, and atomistic modeling of defect formation and transport energetics will provide new, critical understanding about property changes in ZrC. The behavior of materials under irradiation is determined by the balance between damage production, defect clustering, and lattice response. In order to predict those effects at high temperatures so targeted testing can be expanded and extrapolated beyond the known database, it is necessary to determine the defect energetics and mobilities as these control damage accumulation and annealing. In particular, low-temperature irradiations are invaluable for determining the regions of defect mobility. Computer simulation techniques are particularly useful for identifying basic defect properties, especially if closely coupled with a well-constructed and complete experimental database. The close coupling of calculation and experiment in this project will provide mutual benchmarking and allow us to glean a deeper understanding of the irradiation response of ZrC, which can then be applied to the prediction of its behavior in reactor conditions.
Author: Igor L. Shabalin
Publisher: Springer
ISBN: 9402413022
Category : Technology & Engineering
Languages : en
Pages : 755
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This exhaustive work in three volumes and over 1300 pages provides a thorough treatment of ultra-high temperature materials with melting points over 2500 °C. The first volume focuses on Carbon and Refractory Metals, whilst the second and third are dedicated solely to Refractory compounds and the third to Refractory Alloys and Composites respectively. Topics included are physical (crystallographic, thermodynamic, thermo physical, electrical, optical, physico-mechanical, nuclear) and chemical (solid-state diffusion, interaction with chemical elements and compounds, interaction with gases, vapours and aqueous solutions) properties of the individual physico-chemical phases of carbon (graphite/graphene), refractory metals (W, Re, Os, Ta, Mo, Nb, Ir) and compounds (oxides, nitrides, carbides, borides, silicides) with melting points in this range. It will be of interest to researchers, engineers, postgraduate, graduate and undergraduate students alike. The reader is provided with the full qualitative and quantitative assessment for the materials, which could be applied in various engineering devices and environmental conditions at ultra-high temperatures, on the basis of the latest updates in the field of physics, chemistry, materials science and engineering.
Author: M. Griffiths
Publisher:
ISBN:
Category :
Languages : en
Pages : 0
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Author: Kimberly Colas
Publisher:
ISBN:
Category : Zirconium oxide
Languages : en
Pages : 30
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Book Description
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.
Author: M. Griffiths
Publisher:
ISBN:
Category :
Languages : en
Pages : 0
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Author: GARY S. WAS
Publisher: Springer
ISBN: 1493934384
Category : Technology & Engineering
Languages : en
Pages : 1014
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Book Description
The revised second edition of this established text offers readers a significantly expanded introduction to the effects of radiation on metals and alloys. It describes the various processes that occur when energetic particles strike a solid, inducing changes to the physical and mechanical properties of the material. Specifically it covers particle interaction with the metals and alloys used in nuclear reactor cores and hence subject to intense radiation fields. It describes the basics of particle-atom interaction for a range of particle types, the amount and spatial extent of the resulting radiation damage, the physical effects of irradiation and the changes in mechanical behavior of irradiated metals and alloys. Updated throughout, some major enhancements for the new edition include improved treatment of low- and intermediate-energy elastic collisions and stopping power, expanded sections on molecular dynamics and kinetic Monte Carlo methodologies describing collision cascade evolution, new treatment of the multi-frequency model of diffusion, numerous examples of RIS in austenitic and ferritic-martensitic alloys, expanded treatment of in-cascade defect clustering, cluster evolution, and cluster mobility, new discussion of void behavior near grain boundaries, a new section on ion beam assisted deposition, and reorganization of hardening, creep and fracture of irradiated materials (Chaps 12-14) to provide a smoother and more integrated transition between the topics. The book also contains two new chapters. Chapter 15 focuses on the fundamentals of corrosion and stress corrosion cracking, covering forms of corrosion, corrosion thermodynamics, corrosion kinetics, polarization theory, passivity, crevice corrosion, and stress corrosion cracking. Chapter 16 extends this treatment and considers the effects of irradiation on corrosion and environmentally assisted corrosion, including the effects of irradiation on water chemistry and the mechanisms of irradiation-induced stress corrosion cracking. The book maintains the previous style, concepts are developed systematically and quantitatively, supported by worked examples, references for further reading and end-of-chapter problem sets. Aimed primarily at students of materials sciences and nuclear engineering, the book will also provide a valuable resource for academic and industrial research professionals. Reviews of the first edition: "...nomenclature, problems and separate bibliography at the end of each chapter allow to the reader to reach a straightforward understanding of the subject, part by part. ... this book is very pleasant to read, well documented and can be seen as a very good introduction to the effects of irradiation on matter, or as a good references compilation for experimented readers." - Pauly Nicolas, Physicalia Magazine, Vol. 30 (1), 2008 “The text provides enough fundamental material to explain the science and theory behind radiation effects in solids, but is also written at a high enough level to be useful for professional scientists. Its organization suits a graduate level materials or nuclear science course... the text was written by a noted expert and active researcher in the field of radiation effects in metals, the selection and organization of the material is excellent... may well become a necessary reference for graduate students and researchers in radiation materials science.” - L.M. Dougherty, 07/11/2008, JOM, the Member Journal of The Minerals, Metals and Materials Society.
Author: Todd R. Allen
Publisher: ASTM International
ISBN: 0803134010
Category : Materials
Languages : en
Pages : 411
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