Download or read book STATUS OF HIGH FLUX ISOTOPE REACTOR IRRADIATION OF SILICON CARBIDE SILICON CARBIDE JOINTS written by and published by . This book was released on 2014 with total page pages. Available in PDF, EPUB and Kindle. Book excerpt: Development of silicon carbide (SiC) joints that retain adequate structural and functional properties in the anticipated service conditions is a critical milestone toward establishment of advanced SiC composite technology for the accident-tolerant light water reactor (LWR) fuels and core structures. Neutron irradiation is among the most critical factors that define the harsh service condition of LWR fuel during the normal operation. The overarching goal of the present joining and irradiation studies is to establish technologies for joining SiC-based materials for use as the LWR fuel cladding. The purpose of this work is to fabricate SiC joint specimens, characterize those joints in an unirradiated condition, and prepare rabbit capsules for neutron irradiation study on the fabricated specimens in the High Flux Isotope Reactor (HFIR). Torsional shear test specimens of chemically vapor-deposited SiC were prepared by seven different joining methods either at Oak Ridge National Laboratory or by industrial partners. The joint test specimens were characterized for shear strength and microstructures in an unirradiated condition. Rabbit irradiation capsules were designed and fabricated for neutron irradiation of these joint specimens at an LWR-relevant temperature. These rabbit capsules, already started irradiation in HFIR, are scheduled to complete irradiation to an LWR-relevant dose level in early 2015.

Download or read book Radiation Effects in Silicon Carbide written by A.A. Lebedev and published by Materials Research Forum LLC. This book was released on 2017 with total page 172 pages. Available in PDF, EPUB and Kindle. Book excerpt: The book reviews the most interesting research concerning the radiation defects formed in 6H-, 4H-, and 3C-SiC under irradiation with electrons, neutrons, and some kinds of ions. The electrical parameters that make SiC a promising material for applications in modern electronics are discussed in detail. Specific features of the crystal structure of SiC are considered. It is shown that, when wide-bandgap semiconductors are studied, it is necessary to take into account the temperature dependence of the carrier removal rate, which is a standard parameter for determining the radiation hardness of semiconductors. The carrier removal rate values obtained by irradiation of various SiC polytypes with n- and p-type conductivity are analyzed in relation to the type and energy of the irradiating particles. The influence exerted by the energy of charged particles on how radiation defects are formed and conductivity is compensated in semiconductors under irradiation is analyzed. Furthermore, the possibility to produce controlled transformation of silicon carbide polytype is considered. The involvement of radiation defects in radiative and nonradiative recombination processes in SiC is analyzed. Data are also presented regarding the degradation of particular SiC electronic devices under the influence of radiation and a conclusion is made regarding the radiation resistance of SiC. Lastly, the radiation hardness of devices based on silicon and silicon carbide are compared.

Download or read book The Effect of Grain Size on the Radiation Response of Silicon Carbide and Its Dependence on Irradiation Species and Temperature written by and published by . This book was released on 2014 with total page 0 pages. Available in PDF, EPUB and Kindle. Book excerpt: In recent years the push for green energy sources has intensified, and as part of that effort accident tolerant and more efficient nuclear reactors have been designed. These reactors demand exceptional material performance, as they call for higher temperatures and doses. Silicon carbide (SiC) is a strong candidate material for many of these designs due to its low neutron cross-section, chemical stability, and high temperature resistance. The possibility of improving the radiation resistance of SiC by reducing the grain size (thus increasing the sink density) is explored in this work. In-situ electron irradiation and Kr ion irradiation was utilized to explore the radiation resistance of nanocrystalline SiC (nc-SiC), SiC nanopowders, and microcrystalline SiC. Electron irradiation simplifies the experimental results, as only isolated Frenkel pairs are produced so any observed differences are simply due to point defect interactions with the original microstructure. Kr ion irradiation simulates neutron damage, as large radiation cascades with a high concentration of point defects are produced. Kr irradiation studies found that radiation resistance decreased with particle size reduction and grain refinement (comparing nc-SiC and microcrystalline SiC). This suggests that an interface-dependent amorphization mechanism is active in SiC, suggested to be interstitial starvation. However, under electron irradiation it was found that nc-SiC had improved radiation resistance compared to single crystal SiC. This was found to be due to several factors including increased sink density and strength and the presence of stacking faults. The stacking faults were found to improve radiation response by lowering critical energy barriers. The change in radiation response between the electron and Kr ion irradiations is hypothesized to be due to either the change in ion type (potential change in amorphization mechanism) or a change in temperature (at the higher temperatures of the Kr ion irradiation, critical energy barriers can be overcome without the assistance of stacking faults). The dependence of the radiation response of SiC on grain size is not as straight forward as initially presumed. The stacking faults present in many nc-SiC materials boost radiation resistance, but an increased number of interfaces may lead to a reduction in radiation response.

Download or read book The Effect of Irradiation on Siliconized silicon Carbide Coatings for Graphite written by J. L. Jackson and published by . This book was released on 1961 with total page 36 pages. Available in PDF, EPUB and Kindle. Book excerpt:

Download or read book Microstructure of Commercially Available Silicon Carbide Following High Energy Neutron and Dual Ion Irradiation written by Steven D. Harrison and published by . This book was released on 1984 with total page 118 pages. Available in PDF, EPUB and Kindle. Book excerpt:

Download or read book Fast Reactor Irradiation of Pyrolytic Silicon Carbide written by B.E. Sheldon and published by . This book was released on 1975 with total page 17 pages. Available in PDF, EPUB and Kindle. Book excerpt:

Download or read book Stored Energy in Irradiated Silicon Carbide written by and published by . This book was released on 1997 with total page 6 pages. Available in PDF, EPUB and Kindle. Book excerpt: This report presents a short review of the phenomenon of Wigner stored energy release from irradiated graphite and discusses it in relation to neutron irradiation of silicon carbide. A single published work in the area of stored energy release in SiC is reviewed and the results are discussed. It appears from this previous work that because the combination of the comparatively high specific heat of SiC and distribution in activation energies for recombining defects, the stored energy release of SiC should only be a problem at temperatures lower than those considered for fusion devices. The conclusion of this preliminary review is that the stored energy release in SiC will not be sufficient to cause catastrophic heating in fusion reactor components, though further study would be desirable.

Download or read book Silicon Carbide Oxidation in High Temperature Steam written by Ramsey Paul Arnold and published by . This book was released on 2011 with total page 126 pages. Available in PDF, EPUB and Kindle. Book excerpt: The commercial nuclear power industry is continually looking for ways to improve reactor productivity and efficiency and to increase reactor safety. A concern that is closely regulated by the Nuclear Regulatory Commission is the exothermic zircaloy-steam oxidation reaction which can potentially occur during a loss of coolant accident (LOCA), and may become autocatalytic beyond 1,200 0C, thus generating a large amount of hydrogen. The concern for the zircaloy oxidation reaction has been heightened since the March 2011 events of Fukushima, Japan. One solution offering promising results is the use of silicon carbide (SiC) cladding in nuclear reactor fuel rod designs. SiC, a robust ceramic which reacts very slowly with water or steam, has many features that meet or exceed that of zircaloy including the ability to withstand higher temperatures due to a higher melting point and the ability to absorb fewer neutrons than zircaloy which would allow for increased safety margins and fuel burnup. An experimental investigation of the oxidation performance of a-SiC during a postulated LOCA event was performed. The test facility was designed and fabricated to test the oxidation rates of zircaloy and SiC in a high temperature, high-purity, flowing steam environment. Studies of zircaloy-4 oxidation were conducted to validate the test facility for this purpose. Thirty six zircaloy-4 tests lasting up to 30 minutes, at temperatures ranging from 800°C to 1,200°C, were completed and compared to existing models and literature data. Additionally, six longer duration a-SiC tests lasting from 8 hours to 48 hours, at temperatures of 1,140°C and 1,200°C, were completed. These tests clearly show that, from an oxidation perspective, SiC significantly outperforms zircaloy in high-flowing, superheated steam. For zircaloy, results from the most intense temperature/duration testing combination of 1,200°C for 30 minutes show 15.6 percent weight gain. For the most intense SiC tests at 1,200°C for eight hours, a weight loss of two orders of magnitude less occurred, a 0.077 percent weight loss. The four 24 hour and 48 hour SiC tests at 1,140°C also correlate well with the expected paralinear oxidation trend and further confirm that SiC is more resistant to oxidation in high temperature steam than zircaloy.

Download or read book Nuclear Science Abstracts written by and published by . This book was released on 1976 with total page 1026 pages. Available in PDF, EPUB and Kindle. Book excerpt:

Download or read book Ceramic Matrix Composites written by Narottam P. Bansal and published by John Wiley & Sons. This book was released on 2014-10-27 with total page 725 pages. Available in PDF, EPUB and Kindle. Book excerpt: This book is a comprehensive source of information on various aspects of ceramic matrix composites (CMC). It covers ceramic and carbon fibers; the fiber-matrix interface; processing, properties and industrial applications of various CMC systems; architecture, mechanical behavior at room and elevated temperatures, environmental effects and protective coatings, foreign object damage, modeling, life prediction, integration and joining. Each chapter in the book is written by specialists and internationally renowned researchers in the field. This book will provide state-of-the-art information on different aspects of CMCs. The book will be directed to researchers working in industry, academia, and national laboratories with interest and professional competence on CMCs. The book will also be useful to senior year and graduate students pursuing degrees in ceramic science and engineering, materials science and engineering, aeronautical, mechanical, and civil or aerospace engineering. Presents recent advances, new approaches and discusses new issues in the field, such as foreign object damage, life predictions, multiscale modeling based on probabilistic approaches, etc. Caters to the increasing interest in the application of ceramic matrix composites (CMC) materials in areas as diverse as aerospace, transport, energy, nuclear, and environment. CMCs are considered ans enabling technology for advanced aeropropulsion, space propulsion, space power, aerospace vehicles, space structures, as well as nuclear and chemical industries. Offers detailed descriptions of ceramic and carbon fibers; fiber-matrix interface; processing, properties and industrial applications of various CMC systems; architecture, mechanical behavior at room and elevated temperatures, environmental effects and protective coatings, foreign object damage, modeling, life prediction, integration/joining.

Download or read book Threshold Irradiation Dose for Amorphization of Silicon Carbide written by and published by . This book was released on 1997 with total page 13 pages. Available in PDF, EPUB and Kindle. Book excerpt: The amorphization of silicon carbide due to ion and electron irradiation is reviewed with emphasis on the temperature-dependent critical dose for amorphization. The effect of ion mass and energy on the threshold dose for amorphization is summarized, showing only a weak dependence near room temperature. Results are presented for 0.56 MeV silicon ions implanted into single crystal 6H-SiC as a function of temperature and ion dose. From this, the critical dose for amorphization is found as a function of temperature at depths well separated from the implanted ion region. Results are compared with published data generated using electrons and xenon ions as the irradiating species. High resolution TEM analysis is presented for the Si ion series showing the evolution of elongated amorphous islands oriented such that their major axis is parallel to the free surface. This suggests that surface or strain effects may be influencing the apparent amorphization threshold. Finally, a model for the temperature threshold for amorphization is described using the Si ion irradiation flux and the fitted interstitial migration energy which was found to be (approximately)0.56eV. This model successfully explains the difference in the temperature dependent amorphization behavior of SiC irradiated with 0.56 MeV Si at 1 x 10−3 dpa/s and with fission neutrons irradiated at 1 x 10−6 dpa/s irradiated to 15 dpa in the temperature range of (approximately)340"10K.

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 Silicon Carbide Performance as Cladding for Advanced Uranium and Thorium Fuels for Light Water Reactors written by Yanin Sukjai and published by . This book was released on 2014 with total page 341 pages. Available in PDF, EPUB and Kindle. Book excerpt: There has been an ongoing interest in replacing the fuel cladding zirconium-based alloys by other materials to reduce if not eliminate the autocatalytic and exothermic chemical reaction with water and steam at above 1,200 °C. The search for an accident tolerant cladding intensified after the Fukushima events of 2011. Silicon carbide (SiC) possesses several desirable characteristics as fuel cladding in light water reactors (LWRs). Compared to zirconium, SiC has higher melting point, higher strength at elevated temperature, and better dimensional stability when exposed to radiation, as well as lower thermal expansion, creep rate, and neutron absorption cross-section. However, under irradiation, the thermal conductivity of SiC is degraded considerably. Furthermore, lack of creep down towards the fuel causes the fuel-cladding gap and gap thermal resistance to stay relatively large during in-core service. This leads to higher fuel temperature during irradiation. In order to reduce the high fuel temperature during operation, the following fuel design options were investigated in this study: using beryllium oxide (BeO) additive to enhance fuel thermal conductivity, changing the gap bond material from helium to lead-bismuth eutectic (LBE) and adding a central void in the fuel pellet. In addition, the consequences of using thorium oxide (ThO2) as host matrix for plutonium oxide (PuO2) were covered. The effects of cladding thickness on fuel performance were also analyzed. A steady-state fuel performance modeling code, FRAPCON 3.4, was used as a primary tool in this study. Since the official version of the code does not include the options mentioned above, modifications of the source code were necessary. All of these options have been modeled and integrated into a single version of the code called FRAPCON 3.4-MIT. Moreover, material properties including thermal conductivity, swelling rate, and helium production/release rate of BeO have been updated. Material properties of ThO2 have been added to study performance of ThO2-PuO2 . This modified code was used to study the thermo-mechanical behavior of the most limiting fuel rod with SiC cladding, and explore the possibility to improve the fuel performance with various design options. The fuel rod designs and operating conditions of a 4-loop Westinghouse pressurized water reactors (PWR) and Babcock and Wilcox (B&W) mPower small modular reactors (SMR) were reactors (PWR) and Babcock and Wilcox (B&W) mPower small modular reactors (SMR) were chosen as representatives of conventional PWRs and upcoming SMRs, respectively. Sensitivity analyses on initial helium gap pressure, linear heat generation rate (LHGR) history, and peak rod assumptions have been performed. The results suggest that, because of its lower thermal conductivity, SiC is more sensitive to changes in these parameters than zirconium alloys. For a low-conducting material like SiC, an increase in cladding thickness plays a significant role in fuel performance. With a thicker cladding (from 0.57 to 0.89 mm), the temperature drop across the cladding increases, which makes the fuel temperature higher than that with the thin cladding. Reduction of fuel volume to accommodate the thicker cladding also causes negative impact on fuel performance. However, if the extra volume of the cladding replaces some coolant, the reduced coolant fraction design (RCF) has superior performance to the decreased fuel volume fraction design. In general, the most effective fuel temperature improvement option appears to be the option of mixing beryllium oxide into the fuel. This method outperforms others because it improves the overall thermal conductivity and reduces the overall temperature of the fuel. With lower fuel temperature, fission gas release and eventually plenum pressure -- one of the most life-limiting factor for SiC -- can be lowered.

Download or read book Irradiation damage in reaction bonded silicon carbide written by R. B. Matthews and published by . This book was released on 1974 with total page 0 pages. Available in PDF, EPUB and Kindle. Book excerpt:

Download or read book Irradiation Behavior of Pyrolytic Silicon Carbide HTGR written by and published by . This book was released on 1983 with total page pages. Available in PDF, EPUB and Kindle. Book excerpt: Fuel particles for the High-Temperature Gas-Cooled Reactor (HTGR) contain a layer of pyrolytic silicon carbide to act as a miniature pressure vessel and primary fission-product barrier. Optimization of the SiC with respect to fuel performance involves four areas of study: characterization of as-deposited SiC coatings; thermodynamics and kinetics of chemical reactions between SiC and fission products; irradiation behavior of SiC in the absence of fission products; and combined effects of irradiation and fission products. This paper reports the behavior of SiC deposited on inert microspheres and irradiated to fast-neutron fluences typical of HTGR fuel at end-of-life.

Download or read book Silicon Carbide Temperature Monitor Measurements at the High Temperature Test Laboratory written by and published by . This book was released on 2010 with total page pages. Available in PDF, EPUB and Kindle. Book excerpt: Silicon carbide (SiC) temperature monitors are now available for use as temperature sensors in Advanced Test Reactor (ATR) irradiation test capsules. Melt wires or paint spots, which are typically used as temperature sensors in ATR static capsules, are limited in that they can only detect whether a single temperature is or is not exceeded. SiC monitors are advantageous because a single monitor can be used to detect for a range of temperatures that may have occurred during irradiation. As part of the efforts initiated by the ATR National Scientific User Facility (NSUF) to make SiC temperature monitors available, a capability was developed to complete post-irradiation evaluations of these monitors. As discussed in this report, the Idaho National Laboratory (INL) selected the resistance measurement approach for detecting peak irradiation temperature from SiC temperature monitors. This document describes the INL efforts to develop the capability to complete these resistance measurements. In addition, the procedure is reported that was developed to assure that high quality measurements are made in a consistent fashion.

Download or read book Atomic Resolution Investigation of Irradiation Induced Defects in Silicon Carbide written by and published by . This book was released on 2014 with total page pages. Available in PDF, EPUB and Kindle. Book excerpt: