Microstructural Evolution of Zirconium Carbide (ZrCx) Ceramics Under Irradiation Conditions PDF Download
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Author: Raul Fernando Florez Meza Florez
Publisher:
ISBN:
Category :
Languages : en
Pages : 179
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Book Description
"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: Raul Fernando Florez Meza Florez
Publisher:
ISBN:
Category :
Languages : en
Pages : 179
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Book Description
"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: 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:
Publisher:
ISBN:
Category :
Languages : en
Pages :
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Book Description
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: J. Gan
Publisher:
ISBN:
Category : Dislocation
Languages : en
Pages : 7
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Book Description
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:
Publisher: Elsevier
ISBN: 0081028660
Category : Science
Languages : en
Pages : 4871
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Book Description
Materials in a nuclear environment are exposed to extreme conditions of radiation, temperature and/or corrosion, and in many cases the combination of these makes the material behavior very different from conventional materials. This is evident for the four major technological challenges the nuclear technology domain is facing currently: (i) long-term operation of existing Generation II nuclear power plants, (ii) the design of the next generation reactors (Generation IV), (iii) the construction of the ITER fusion reactor in Cadarache (France), (iv) and the intermediate and final disposal of nuclear waste. In order to address these challenges, engineers and designers need to know the properties of a wide variety of materials under these conditions and to understand the underlying processes affecting changes in their behavior, in order to assess their performance and to determine the limits of operation. Comprehensive Nuclear Materials, Second Edition, Seven Volume Set provides broad ranging, validated summaries of all the major topics in the field of nuclear material research for fission as well as fusion reactor systems. Attention is given to the fundamental scientific aspects of nuclear materials: fuel and structural materials for fission reactors, waste materials, and materials for fusion reactors. The articles are written at a level that allows undergraduate students to understand the material, while providing active researchers with a ready reference resource of information. Most of the chapters from the first Edition have been revised and updated and a significant number of new topics are covered in completely new material. During the ten years between the two editions, the challenge for applications of nuclear materials has been significantly impacted by world events, public awareness, and technological innovation. Materials play a key role as enablers of new technologies, and we trust that this new edition of Comprehensive Nuclear Materials has captured the key recent developments. Critically reviews the major classes and functions of materials, supporting the selection, assessment, validation and engineering of materials in extreme nuclear environments Comprehensive resource for up-to-date and authoritative information which is not always available elsewhere, even in journals Provides an in-depth treatment of materials modeling and simulation, with a specific focus on nuclear issues Serves as an excellent entry point for students and researchers new to the field
Author: M. Griffiths
Publisher:
ISBN:
Category : Climb
Languages : en
Pages : 8
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Book Description
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.
Author: M. Griffiths
Publisher:
ISBN:
Category :
Languages : en
Pages : 0
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Author: M. Griffiths
Publisher:
ISBN:
Category :
Languages : en
Pages : 0
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Author: YASIR. Idrees
Publisher:
ISBN:
Category :
Languages : en
Pages :
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Author: M. Griffiths
Publisher:
ISBN:
Category : Congress
Languages : en
Pages : 23
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Book Description
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.
