Download or read book Atomistic Modelling of Irradiation induced Microstructure Evolution in Fe Alloys written by Ebrahim Mansouri and published by . This book was released on 2024 with total page 0 pages. Available in PDF, EPUB and Kindle. Book excerpt:

Download or read book Microstructure Evolution During Irradiation Volume 439 written by Ian M. Robertson and published by . This book was released on 1997-06-25 with total page 770 pages. Available in PDF, EPUB and Kindle. Book excerpt: This book from MRS discusses the evolution of a material's microstructure as a result of its interaction with energetic particles such as ions, neutrons or electrons. The book is inter-disciplinary and emphasizes all classes of materials including metals, intermetallic compounds, ceramics, polymers, superconductors, semiconductors and insulators. A strong focus is placed on experimental techniques for measuring and quantifying damage and microstructure changes, and on computer simulation techniques for predicting and understanding this phenomena. Topics include: ion-implantation damage in semiconductors; radiation damage in metals; radiation damage in ceramics; radiation effects in polymers and beam-induced effects.

Download or read book Modelling of Radiation Induced Segregation in Austenitic Fe Alloys at the Atomistic Level written by Jean-Baptiste Piochaud and published by . This book was released on 2013 with total page 0 pages. Available in PDF, EPUB and Kindle. Book excerpt: In pressurized water reactors, under irradiation internal structures are subject of irradiation assisted stress corrosion cracking which is influenced by radiation induced segregation (RIS). In this work RIS of 316 stainless steels is modelled considering a model ternary Fe-10Ni-20Cr alloy. For this purpose we have built an Fe-Ni-Cr pair interaction model to simulate RIS at the atomistic level using an atomistic kinetic Monte Carlo approach. The pair interactions have been deduced from density functional theory (DFT) data available in the pure fcc systems but also from DFT calculations we have performed in the Fe-10Ni-20Cr target alloy. Point defect formation energies were calculated and found to depend strongly on the local environment of the defect. As a consequence, a rather good estimation of these energies can be obtained from the knowledge of the number and respective positions of the Ni and Cr atoms in the vicinity of the defect. This work shows that a model based only on interaction parameters between elements positioned in perfect lattice sites (solute atoms and vacancy) cannot capture alone both the thermodynamic and the kinetic aspect of RIS. A more accurate of estimating the barriers encountered by the diffusing species is required than the one used in our model, which has to depend on the saddle point environment. This study therefore shows thus the need to estimate point defect migration energies using the DFT approach to calibrate a model that can be used in the framework of atomic kinetic Monte Carlo simulations. We also found that the reproduction by our pair interaction model of DFT data for the self-interstitial atoms was found to be incompatible with the modelling of RIS under electron irradiation.

Download or read book Comprehensive Nuclear Materials written by and published by Elsevier. This book was released on 2020-07-22 with total page 4871 pages. Available in PDF, EPUB and Kindle. Book excerpt: 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

Download or read book Investigation of The Synergistic Influence of Irradiation Temperature and Atomic Displacement Rate on the Microstructural Evolution of Ion Irradiated Model Austenitic Alloy Fe 15Cr 16Ni written by and published by . This book was released on 2002 with total page pages. Available in PDF, EPUB and Kindle. Book excerpt: A comprehensive experimental investigation of microstructural evolution has been conducted on Fe-15Cr-16Ni irradiated with 4.0 MeV nickel ions in the High Fluence Irradiation Facility of the University of Tokyo. Irradiations proceeded to dose levels ranging from ~0.2 to ~26 dpa at temperatures of 300, 400 and 500 degrees C at displacement rates of 1 x 10^-4, 4 x 10^-4 and 1 x 10^-3 dpa/sec. This experiment is one of two companion experiments directed toward the study of the dependence of void swelling on displacement rate. The other experiment proceeded at seven different but lower dpa rates in FFTF-MOTA at ~400 degrees C. In both experiments the swelling was found at every irradiation condition studied to monotonically increase with decreases in dpa rate. The microstructural evolution under ion irradiation was found to be very sensitive to the displacement rate at all three temperatures. The earliest and most sensitive component of microstructure to both temperature and especially displacement rate was found to be the Frank loops. The second most sensitive component was found to be the void microstructure, which co-evolves with the loop and dislocation microstructure. These data support the prediction that void swelling will probably be higher in lower-flux fusion devices and PWRs at a given irradiation temperature when compared to irradiations conducted at higher dpa rates in fast reactors.

Download or read book Microstructure Evolution in Proton irradiated Austenitic Fe Cr Ni Alloys Under LWR Core Conditions written by Jian Gan and published by . This book was released on 1999 with total page 710 pages. Available in PDF, EPUB and Kindle. Book excerpt:

Download or read book Fundamentals of Radiation Materials Science written by GARY S. WAS and published by Springer. This book was released on 2016-07-08 with total page 1014 pages. Available in PDF, EPUB and Kindle. Book excerpt: 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.

Download or read book Modelling of Radiation Damage in Fe Cr Alloys written by L. Malerba and published by . This book was released on 2007 with total page 19 pages. Available in PDF, EPUB and Kindle. Book excerpt: High-Cr ferritic/martensitic steels are being considered as structural materials for a large number of future nuclear applications, from fusion to accelerator-driven systems and GenIV reactors. Fe-Cr alloys can be used as model materials to investigate some of the mechanisms governing their microstructure evolution under irradiation and its correlation to changes in their macroscopic properties. Focusing on these alloys, we show an example of how the integration of computer simulation and theoretical models can provide keys for the interpretation of a host of relevant experimental observations. In particular we show that proper accounting for two basic features of these alloys, namely, the existence of a fairly strong attractive interaction between self-interstitials and Cr atoms and of a mixing enthalpy that changes sign from negative to positive around 8 to 10 % Cr, is a necessary and, to a certain extent, sufficient condition to rationalize and understand their behavior under irradiation. These features have been revealed by ab initio calculations, are supported by experimental evidence, and have been adequately transferred into advanced empirical interatomic potentials, which have been and are being used for the simulation of damage production, defect behavior, and phase transformation in these alloys. The results of the simulations have been and are being used to parameterize models capable of extending the description of radiation effects to scales beyond the reach of molecular dynamics. The present paper intends to highlight the most important achievements and results of this research activity.

Download or read book Irradiation Hardening and Microstructural Evolution in Fe Cu Model Alloys written by M. Narui and published by . This book was released on 2004 with total page 11 pages. Available in PDF, EPUB and Kindle. Book excerpt: Irradiation hardening and microstructural evolution und neutron and electron irradiation have been investigated for pure-Fe and Fe-Cu model alloys. Neutron and electron irradiations were performed in the Japan Material Test Reactor (JMTR) and with using Phodtron electron accelerator at about 290°C and 270±30°C, respectively. Irradiation hardening of pure-Fe and Fe-Cu model alloys is saturated at about 1 x 10-3 dpa in both the neutron and electron irradiation. Irradiation hardening recovered in two temperature ranges. The recovery in the lower temperature range depends on copper concentration and electron irradiation dose, while the recovery at a higher temperature range does not. Recovery behavior of the irradiation hardening suggests indirectly that copper atoms suppress the growth of interstitial clusters. The recovery behavior of positron lifetime does not coincide with that of the hardness, suggesting that the vacancy clusters are not the direct main factor controlling the hardening by matrix damages.

Download or read book Characterization and Modeling of Grain Boundary Chemistry Evolution in Ferritic Steels Under Irradiation written by and published by . This book was released on 2016 with total page 70 pages. Available in PDF, EPUB and Kindle. Book excerpt: Ferritic/martensitic (FM) steels such as HT-9, T-91 and NF12 with chromium concentrations in the range of 9-12 at.% Cr and high Cr ferritic steels (oxide dispersion strengthened steels with 12-18% Cr) are receiving increasing attention for advanced nuclear applications, e.g. cladding and duct materials for sodium fast reactors, pressure vessels in Generation IV reactors and first wall structures in fusion reactors, thanks to their advantages over austenitic alloys. Predicting the behavior of these alloys under radiation is an essential step towards the use of these alloys. Several radiation-induced phenomena need to be taken into account, including phase separation, solute clustering, and radiation-induced segregation or depletion (RIS) to point defect sinks. RIS at grain boundaries has raised significant interest because of its role in irradiation assisted stress corrosion cracking (IASCC) and corrosion of structural materials. Numerous observations of RIS have been reported on austenitic stainless steels where it is generally found that Cr depletes at grain boundaries, consistently with Cr atoms being oversized in the fcc Fe matrix. While FM and ferritic steels are also subject to RIS at grain boundaries, unlike austenitic steels, the behavior of Cr is less clear with significant scatter and no clear dependency on irradiation condition or alloy type. In addition to the lack of conclusive experimental evidence regarding RIS in F-M alloys, there have been relatively few efforts at modeling RIS behavior in these alloys. The need for predictability of materials behavior and mitigation routes for IASCC requires elucidating the origin of the variable Cr behavior. A systematic detailed high-resolution structural and chemical characterization approach was applied to ion-implanted and neutron-irradiated model Fe-Cr alloys containing from 3 to 18 at.% Cr. Atom probe tomography analyses of the microstructures revealed slight Cr clustering and segregation to dislocations and grain boundaries in the ion-irradiated alloys. More significant segregation was observed in the neutron irradiated alloys. For the more concentrated alloys, irradiation did not affect existing Cr segregation to grain boundaries and segregation to dislocation loops was not observed perhaps due to a change in the dislocation loop structure with increasing Cr concentration. Precipitation of [alpha]' was observed in the neutron irradiated alloys containing over 9 at.% Cr. However ion irradiation appears to suppress the precipitation process. Initial low dose ion irradiation experiments strongly suggest a cascade recoil effect. The systematic analysis of grain boundary orientation on Cr segregation was significantly challenged by carbon contamination during ion irradiation or by existing carbon and therefore carbide formation at grain boundaries (sensitization). The combination of the proposed systematic experimental approach with atomistic modeling of diffusion processes directly addresses the programmatic need for developing and benchmarking predictive models for material degradation taking into account atomistic kinetics parameters.

Download or read book Microstructural Evolution of Ferritic martensitic Steels Under Heavy Ion Irradiation written by Cem Topbasi and published by . This book was released on 2014 with total page pages. Available in PDF, EPUB and Kindle. Book excerpt: Ferritic-martensitic steels are primary candidate materials for fuel cladding and internal applications in the Sodium Fast Reactor, as well as first-wall and blanket materials in future fusion concepts because of their favorable mechanical properties and resistance to radiation damage. Since microstructure evolution under irradiation is amongst the key issues for these materials in these applications, developing a fundamental understanding of the irradiation-induced microstructure in these alloys is crucial in modeling and designing new alloys with improved properties.The goal of this project was to investigate the evolution of microstructure of two commercial ferritic-martensitic steels, NF616 and HCM12A, under heavy ion irradiation at a broad temperature range. An in situ heavy ion irradiation technique was used to create irradiation damage in the alloy; while it was being examined in a transmission electron microscope. Electron-transparent samples of NF616 and HCM12A were irradiated in situ at the Intermediate Voltage Electron Microscope (IVEM) at Argonne National Laboratory with 1 MeV Kr ions to ~10 dpa at temperatures ranging from 20 to 773 K. The microstructure evolution of NF616 and HCM12A was followed in situ by systematically recording micrographs and diffraction patterns as well as capturing videos during irradiation.In these irradiations, there was a period during which no changes are visible in the microstructure. After a threshold dose (~0.1 dpa between 20 and 573 K, and ~2.5 dpa at 673 K) black dots started to become visible under the ion beam. These black dots appeared suddenly (from one frame to the next) and are thought to be small defect clusters (2-5 nm in diameter), possibly small dislocation loops with Burgers vectors of either 1/2111 or 100.The overall density of these defect clusters increased with dose and saturated around 6 dpa. At saturation, a steady-state is reached in which defects are eliminated and created at the same rates so that the defect density is constant. After saturation, defects constantly appeared and disappeared in a time that is shorter than the time in between frames (normally 34 ms). The average diameter and size distribution of the irradiation-induced defect clusters did not vary with dose during a single irradiation in the temperature range of 50 to 573 K in NF616, and 20 to 673 K in HCM12A. At 673 K, the defects in NF616 grew and coalesced under irradiation which led to larger average defect sizes and low defect density. At high doses extended defect structures in NF616 formed as short segments aligned along 100 directions. At 773 K, the frequency of defect formation per unit area was the lowest amongst all irradiations and all the visible defect clusters that formed eventually faded out gradually (in ~28 seconds) leading to no net defect accumulation in NF616 even at the highest irradiation dose of 10 dpa.Under irradiation, a significant fraction of these defect clusters exhibited sudden one-dimensional jumps (over ~5nm) between 20 and 573 K, that is, some defect clusters move "or jump" along 211 directions which is consistent with the expected Burgers vector direction of (111). Interestingly, at 673 and 773 K, defects in NF616 and HCM12A did not exhibit the sudden jumps and jerks that were frequently observed during lower temperature irradiations. No resolvable loops, voids or precipitates were formed in NF616 and HCM12A. Furthermore, no significant interaction of the irradiation induced defects with the foil surface, pre-existing dislocation network or grain boundaries was observed between 20 and 773 K.A simplified rate theory model was developed to describe the initial defect formation processes. The model is based on the reactions between intra-cascade clusters driven by the one-dimensional movement of sub-visible interstitial clusters in their glide cylinder under irradiation after detrapping from interstitial and substitutional solute atoms by cascade impact. Multiple cascade impacts on previously existing clusters allow them to gather clusters during their glide, leading to the formation of TEM-visible (~2 nm) defects. The low dose defect density approximated by model is in good agreement with the experimental results. In addition, the model rationalizes the threshold dose before which no visible defect clusters were formed.

Download or read book Radiation Tolerance of Neutron Irradiated Model Fe Cr Al Alloys written by and published by . This book was released on 2015 with total page pages. Available in PDF, EPUB and Kindle. Book excerpt: The Fe Cr Al alloy system has the potential to form an important class of enhanced accident-tolerant cladding materials in the nuclear power industry owing to the alloy system's higher oxidation resistance in high-temperature steam environments compared with traditional zirconium-based alloys. However, radiation tolerance of Fe Cr Al alloys has not been fully established. In this study, a series of Fe Cr Al alloys with 10 18 wt % Cr and 2.9 4.9 wt % Al were neutron irradiated at 382 C to 1.8 dpa to investigate the irradiation-induced microstructural and mechanical property evolution as a function of alloy composition. Dislocation loops with Burgers vector of a/2 111 and a 100 were detected and quantified. Results indicate precipitation of Cr-rich is primarily dependent on the bulk chromium composition. Mechanical testing of sub-size-irradiated tensile specimens indicates the hardening response seen after irradiation is dependent on the bulk chromium composition. A structure property relationship was developed; it indicated that the change in yield strength after irradiation is caused by the formation of these radiation-induced defects and is dominated by the large number density of Cr-rich precipitates at sufficiently high chromium contents after irradiation.

Download or read book Atomic Scale Modeling of Defect Production and Microstructure Evolution in Irradiated Metals written by and published by . This book was released on 1997 with total page 7 pages. Available in PDF, EPUB and Kindle. Book excerpt: Irradiation effects in materials depend in a complex way on the form of the as-produced primary damage state and its spatial and temporal evolution. Thus, while collision cascades produce defects on a time scale of tens of picosecond, diffusion occurs over much longer time scales, of the order of seconds, and microstructure evolution over even longer time scales. In this report the authors present work aimed at describing damage production and evolution in metals across all the relevant time and length scales. They discuss results of molecular dynamics simulations of displacement cascades in Fe and V. They show that interstitial clusters are produced in cascades above 5 keV, but not vacancy clusters. Next, they discuss the development of a kinetic Monte Carlo model that enables calculations of damage evolution over much longer time scales (1000`s of s) than the picosecond lifetime of the cascade. They demonstrate the applicability of the method by presenting predictions on the fraction of freely migrating defects in [alpha]Fe during irradiation at 600 K.

Download or read book Irradiation Induced Microstructural Evolution and Mechanical Properties in Iron With and Without Helium written by and published by . This book was released on 2008 with total page pages. Available in PDF, EPUB and Kindle. Book excerpt:

Download or read book Energy Research Abstracts written by and published by . This book was released on 1993 with total page 762 pages. Available in PDF, EPUB and Kindle. Book excerpt: Semiannual, with semiannual and annual indexes. References to all scientific and technical literature coming from DOE, its laboratories, energy centers, and contractors. Includes all works deriving from DOE, other related government-sponsored information, and foreign nonnuclear information. Arranged under 39 categories, e.g., Biomedical sciences, basic studies; Biomedical sciences, applied studies; Health and safety; and Fusion energy. Entry gives bibliographical information and abstract. Corporate, author, subject, report number indexes.

Download or read book Proceedings of the 2023 Water Reactor Fuel Performance Meeting written by Jianqiao Liu and published by Springer Nature. This book was released on 2023-11-30 with total page 384 pages. Available in PDF, EPUB and Kindle. Book excerpt: The Water Reactor Fuel Performance Meeting (WRFPM) held in Asia has merged with TopFuel in Europe and LWR Fuel Performance in the United States to form the globally most influential conference in the field of nuclear fuel research. WRFPM2023 is organized by Chinese Nuclear Society (CNS) in cooperation with the Atomic Energy Society of Japan (AESJ), Korean Nuclear Society (KNS), European Nuclear Society (ENS), American Nuclear Society (ANS), the Interna-tional Atomic Energy Agency (IAEA) with the support from China Nuclear Energy In¬dustry Corporation (CNEIC) and TVEL. Conference Topics: 1. Advances in water reactor fuel technology and testing 2. Operation and experience 3. Transient and off-normal fuel behaviour and safety related issues 4. Fuel cycle, used fuel storage and transportation 5. Innovative fuel and related issues 6. Fuel modelling, analysis and methodology

Download or read book Modeling and Simulation of Microstructure Evolution in Solidifying Alloys written by Laurentiu Nastac and published by . This book was released on 2004 with total page 285 pages. Available in PDF, EPUB and Kindle. Book excerpt: