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📒Accident Tolerant Materials For Light Water Reactor Fuels ✍ Raul B. Rebak
✏Author : Raul B. Rebak
✏Publisher :
✏Release Date : 2020-01-06
✏Pages : 236
✏ISBN : 9780128175033
✏Available Language : English, Spanish, And French
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✏Accident Tolerant Materials for Light Water Reactor Fuels Book Summary : Accident Tolerant Materials for Light Water Reactor Fuels provides a description of what an accident tolerant fuel is and the benefits and detriments of each concept. The book begins with an introduction to nuclear power as a renewable energy source and the current materials being utilized in light water reactors. It then moves on to discuss the recent advancements being made in accident tolerant fuels, reviewing the specific materials, their fabrication and implementation, environmental resistance, irradiation behavior, and licensing requirements. The book concludes with a look to the future of new power generation technologies. It is written for scientists and engineers working in the nuclear power industry and is the first comprehensive work on this topic. Introduces the fundamental description of accident tolerant fuel, including fabrication and implementation Describes both the benefits and detriments of the various Accident Tolerant Fuel concepts Includes information on the process of materials selection with a discussion of how and why specific materials were chosen, as well as why others failed
📒Thermal Hydraulics Of Accident Tolerant Fuel Concepts And A Preliminary Demonstration Of Casl S Coupled Tools For Bwrs ✍ Jacob Preston Gorton
✏Author : Jacob Preston Gorton
✏Publisher :
✏Release Date : 2018
✏Pages :
✏ISBN : OCLC:1083898035
✏Available Language : English, Spanish, And French
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✏Thermal Hydraulics of Accident Tolerant Fuel Concepts and a Preliminary Demonstration of CASL s Coupled Tools for BWRs Book Summary : Since the 2011 accident at the Daiichi nuclear power plant in Fukushima, Japan, there has been a worldwide effort to develop so-called accident tolerant fuel (ATF) technologies to enhance safety during design basis and beyond design basis accidents. Part of the ATF development effort involves replacing much of the zirconium-based materials in light water reactors (LWRs). This is due to the accelerated oxidation rate of zirconium at high temperatures potentially experienced during severe accidents, which led to the build-up of hydrogen gas and eventual explosions that occurred at the Daiichi nuclear power plant. To be considered as a possible alternative to zirconium, an ATF candidate material must not only have greater oxidation resistance but must also have equal or better performance than zirconium in reactor operations and safety. Two candidate materials that may meet these requirements are iron-chromium-aluminum (FeCrAl) alloys and silicon carbide fiber-reinforced, silicon carbide matrix composites (SiC/SiC). Two studies on ATF concepts are presented in this thesis, which focus on using computer simulations to evaluate the use of FeCrAl as the fuel rod cladding material in a pressurized water reactor (PWR) and the use of SiC/SiC as the fuel assembly channel box material in a boiling water reactor (BWR). Both of these studies are performed using computer modeling, which is one of the first steps for evaluating new design concepts and eventually integrating them into existing reactors. Developing tools that can accurately predict the performance of nuclear reactors with high fidelity is the goal of the Consortium for Advanced Simulation of Light Water Reactors (CASL). Also included in this thesis is a preliminary demonstration of neutronic-to-thermal-hydraulic coupled BWR simulations performed using the CASL tools MPACT and CTF. In the first study, a model of a PWR fuel assembly was created to predict the critical heat flux (CHF) of FeCrAl fuel rod cladding during an imposed 50% overpower condition, which may be representative of an accident condition. CHF is a critical parameter to evaluate for ATF candidate materials because reaching CHF in a fuel rod can cause a rapid increase in temperature in the reactor that may lead to bursting of the cladding and a loss of ability to cool the core. Current correlations used for predicting flow boiling CHF in reactors are not dependent on material or surface characteristics, but this study showed that preliminary pool boiling results could be used to modify existing CHF correlations to make them more applicable to a given material, such as FeCrAl. Preliminary transient flow boiling experiments are also analyzed in this thesis for Inconel 600 and Stainless Steel 316, which pave the way for future flow boiling experiments using FeCrAl. In the second study, BWR fuel assembly models were created with a SiC/SiC channel box to predict a spatial temperature and fast neutron flux distribution in the channel box. The temperature and fast flux distributions were then used as boundary conditions for a finite element model of the channel box created by Oak Ridge National Laboratory to determine the deflection of the channel box due to temperature and neutron flux gradients. It was found in this study that the deflection of the channel box, which was mainly a product of the nonuniform fast flux distribution causing a swelling gradient within the channel box, may lead to interference with control blades in BWR cores. The work presented in this thesis provides new information on two ATF concepts and helps lay the groundwork for future evaluations. Detailed computational evaluations are an important step in the progression and application of these concepts that have the potential to increase the safety of nuclear reactors. The development of high-fidelity computational tools like MPACT/CTF is important for providing accurate simulated results that can be used in advancing the development of ATF concepts.
✏Author : P. Gannon
✏Publisher : The Electrochemical Society
✏Release Date : 2018-05-04
✏Pages : 93
✏ISBN : 9781607688303
✏Available Language : English, Spanish, And French
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✏High Temperature Corrosion and Materials Chemistry 13 Book Summary :
📒Fundamentals Of Nuclear Engineering ✍ Brent J. Lewis
✏Author : Brent J. Lewis
✏Publisher : John Wiley & Sons
✏Release Date : 2017-06-19
✏Pages : 984
✏ISBN : 9781119271499
✏Available Language : English, Spanish, And French
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✏Fundamentals of Nuclear Engineering Book Summary : Fundamental of Nuclear Engineering is derived from over 25 years of teaching undergraduate and graduate courses on nuclear engineering. The material has been extensively class tested and provides the most comprehensive textbook and reference on the fundamentals of nuclear engineering. It includes a broad range of important areas in the nuclear engineering field; nuclear and atomic theory; nuclear reactor physics, design, control/dynamics, safety and thermal-hydraulics; nuclear fuel engineering; and health physics/radiation protection. It also includes the latest information that is missing in traditional texts, such as space radiation. The aim of the book is to provide a source for upper level undergraduate and graduate students studying nuclear engineering.
📒High Performance Ceramics X ✍ Wei Pan
✏Author : Wei Pan
✏Publisher : Trans Tech Publications Ltd
✏Release Date : 2018-08-31
✏Pages : 994
✏ISBN : 9783035731767
✏Available Language : English, Spanish, And French
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✏High Performance Ceramics X Book Summary : This book contains the papers presented at the Tenth International Conference on High-Performance Ceramics (CICC-10), which was held from November 4 to 7, 2017, in Nanchang, China. The Collection covers most aspects in the field of high-performance ceramics (advanced ceramics), including processing, microstructures, and properties of structural and functional ceramics. Especially, the papers gave an overview of the most recent development in high-performance ceramics in China.
📒Impact Of Reactor Environment On Quenching Heat Transfer Of Accident Tolerant Fuel Cladding ✍ Arunkumar Seshadri
✏Author : Arunkumar Seshadri
✏Publisher :
✏Release Date : 2018
✏Pages : 123
✏ISBN : OCLC:1103919414
✏Available Language : English, Spanish, And French
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✏Impact of Reactor Environment on Quenching Heat Transfer of Accident Tolerant Fuel Cladding Book Summary : Development of accident tolerant fuels (ATF) for light water reactors (LWRs) came into focus for the nuclear engineering community after the accidents at Fukushima-Daiichi. The primary focus of the ATF program is to identify alternative fuel and cladding technologies that may provide enhanced safety, competitiveness, and economics. The new fuel design must also be compatible with present-day LWR design. For near-term applications, coatings on the nominal Zirconium-based cladding material and other metallic materials are being considered to improve the corrosion resistance and reduce the generation of hydrogen at high temperatures. Major ATF coating choices under consideration include chromium as a coating, iron-chromium-aluminum alloys (FeCrAl) as cladding and molybdenum as a coating, which have demonstrated better mechanical and oxidation behavior during the experimental testing. Thermal-fluids characteristics are pivotal for a robust testing of ATF concepts as the proposed candidates may have an entirely different thermal-hydraulic behavior when compared to Zircaloy-4. ATF coatings may display very different boiling characteristics as a result of different microstructures and surface characteristics. In the present work, transient boiling heat transfer during quenching of the candidate ATF claddings on vertical rodlets is studied experimentally. The candidate ATF material (chromium, FeCrAl, and molybdenum) are applied on Zircaloy-4 rodlets. The vertical solid rodlets are heated to temperatures up to 1000 °C and are quenched in a saturated pool of water at atmospheric pressure. The temperature variation during the quenching of rodlets was recorded insitu with synchronized visualization of boiling regimes over the test specimen using a high-speed video camera. The quench performance of the ATF coatings was analyzed based on the examination of various surface parameters such as wettability, roughness, emissivity and capillary wicking. In order to obtain a more realistic picture of the candidate performance during the emergency cooling reflood phase in a nuclear reactor, the coated rodlets are also oxidized in an autoclave before quenching. The performance of the candidate claddings is evaluated after oxidation and the surface characterized. It was observed from the post-test analysis that the surface characteristics and oxidation had a significant impact on the quench performance of ATF coatings, which varied between different coating materials. In order to better understand the thermal margins in a reactor specific environment, an analysis was performed on samples after exposing them to gamma rays. The gamma rays tend to change the surface wettability through a phenomenon called Radiation Induced Surface Activation. A Gammacell 220E irradiator that uses 12 cobalt-60 pencil sources, arranged axially in a sample chamber at MIT, was used to irradiated the samples. The results of water quenching and contact angle studies showed a higher Leidenfrost temperature and wettability in both samples exposed to gamma irradiation. The detailed microscopic analysis attributed the enhanced wettability to oxidation of the surface under gamma irradiation.
📒Thermo Mechanical Analysis Of Iron Chromium Aluminum Fecral Alloy Cladding For Light Water Reactor Fuel Elements ✍ Ryan Terrence Sweet
✏Author : Ryan Terrence Sweet
✏Publisher :
✏Release Date : 2018
✏Pages : 179
✏ISBN : OCLC:1147976476
✏Available Language : English, Spanish, And French
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✏Thermo mechanical Analysis of Iron chromium aluminum FeCrAl Alloy Cladding for Light Water Reactor Fuel Elements Book Summary : Alternative cladding materials have been proposed to replace the currently used zirconium (Zr)-based alloys, in order to improve the accident tolerance of light water reactor (LWR) fuel. Of these materials, there is a particular focus on iron-chromium-aluminum (FeCrAl) alloys that exhibit much slower oxidation kinetics in high-temperature steam than Zr-alloys. This behavior should decrease the energy release due to oxidation and allow the cladding to remain integral longer in the presence of high temperature steam, making accident mitigation more likely. This dissertation documents efforts to develop fuel performance capabilities to assess the behavior of FeCrAl cladding during normal and transient reactor operating scenarios. Within this work, simulations were performed for FeCrAl cladding using constitutive models and representative reactor operating conditions implemented into the finite-element fuel performance code BISON. Simulations were performed targeting the cladding behavior during normal operation of a boiling water reactor using boundary conditions derived from neutronics data. These simulations indicate that the fuel compliance plays a much larger role in the evolution of the cladding stress state after gap closure for the FeCrAl cladding than for Zircaloy. Individual sensitivity analyses of the fuel and cladding creep responses were then performed, which indicated the influence of compliance for each material, separately, on the stress state of the fuel cladding. To improve calculations of the fuel expansion and compliance, an additional investigation was performed to assess the role of creep, relocation, and explicit fracture in the fuel. Fuel rods using each of these models are simulated under representative conditions and compared to test rod measurements. This analysis provides a start toward the development and incorporation of explicit fracture in fuel performance analysis. Additionally, performance and stability under transient conditions must also be demonstrated for FeCrAl cladding. This analysis focused on modeling the integral thermo-mechanical performance of FeCrAl clad uranium dioxide fuel during transient reactor operation. Results from this simple analysis show similar bursting time and temperature between both FeCrAl and Zircaloy cladding, however, beyond cladding burst in these conditions, the superior high temperature oxidation kinetics of the FeCrAl cladding significantly reduce hydrogen gas production and provide longer fuel integrity.
✏Author : Helena Jin
✏Publisher : Springer Science & Business Media
✏Release Date : 2013-08-30
✏Pages : 387
✏ISBN : 9783319007687
✏Available Language : English, Spanish, And French
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✏Advancement of Optical Methods in Experimental Mechanics Volume 3 Book Summary : Advancement of Optical Methods in Experimental Mechanics: Proceedings of the 2013 Annual Conference on Experimental and Applied Mechanics, the third volume of eight from the Conference, brings together contributions to this important area of research and engineering. The collection presents early findings and case studies on a wide range of optical methods ranging from traditional photoelasticity and interferometry to more recent DIC and DVC techniques, and includes papers in the following general technical research areas: Optical metrology and displacement measurements at different scales Digital holography and experimental mechanics Optical measurement systems using polarized light Surface topology Digital image correlation Optical methods for MEMS and NEMS Three-dimensional imaging and volumetric correlation Imaging methods for thermomechanics applications 3D volumetric flow measurement Applied photoelasticity Optical residual stress measurement techniques Advances in imaging technologies
📒Wettability Of Candidate Accident Tolerant Fuel Atf Cladding Materials In Lwr Conditions ✍ Anupam Jena (S.M.)
✏Author : Anupam Jena (S.M.)
✏Publisher :
✏Release Date : 2020
✏Pages : 70
✏ISBN : OCLC:1191901226
✏Available Language : English, Spanish, And French
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✏Wettability of Candidate Accident Tolerant Fuel ATF Cladding Materials in LWR Conditions Book Summary : Since the 2011 Fukushima accident, substantial research has been dedicated to developing accident tolerant fuel (ATF) cladding materials. These analyses have mostly concentrated on the capability of potential ATF materials to withstand runaway steam oxidation and preserve their mechanical strength and structural integrity under thermal shocks. However, knowledge is relatively deficient for the thermal-hydraulic properties of these materials, particularly under light water reactor (LWR) operating conditions. The surface wettability is particularly important, as it affects the dynamics of the boiling heat transfer process, and consequently, the critical heat flux (CHF) and rewetting temperatures, which are important thermal limits for LWRs. Surface wettability determines nucleation site density, bubble departure diameter, and bubble departure frequency. Therefore, it is essential to quantify the surface wettability of candidate ATF cladding materials to determine their thermal-hydraulic behavior compared to conventional Zircaloy claddings. The surface wettability is usually quantified through the sessile droplet contact angle, which is the angle formed between the liquid-vapor and the liquid-solid interface. The contact angle depends on the fluid, solid, surface finish, and operating conditions, i.e., temperature and pressure. However, most of the measurements available in the literature are performed at low pressure and in an inert atmosphere, which is quite different from the operating conditions of LWRs (i.e., in a steam-saturated atmosphere at a pressure as high as 15.5 MPa or 155 bars). To close this gap, in this study, we designed and built an autoclave-type facility capable of measuring static, advancing, and receding contact angle in steam-saturated atmospheres, from sub-atmospheric conditions up to the critical point of water, i.e., 22.1 MPa (221 bar or 3200 psi) and 374°C. We measured the static contact angle of conventional Zircaloy-4 and candidate ATF cladding materials (e.g., Cr-coated Zr-4, FeCrAl, and SiC). The contact angle decreases with an increase in temperature for all the materials. Rough surfaces showed higher wettability, i.e., lower contact angle, compared to the smooth surfaces. These trends are expected from theory. All the materials showed different wettability under the same temperature and pressure conditions. Individual correlations for temperature dependence for each of them are proposed.
📒Proceedings Of The 18th International Conference On Environmental Degradation Of Materials In Nuclear Power Systems Water Reactors ✍ John H. Jackson
✏Author : John H. Jackson
✏Publisher : Springer
✏Release Date : 2018-12-20
✏Pages : 2532
✏ISBN : 9783030046392
✏Available Language : English, Spanish, And French
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✏Proceedings of the 18th International Conference on Environmental Degradation of Materials in Nuclear Power Systems Water Reactors Book Summary : This two-volume set represents a collection of papers presented at the 18th International Conference on Environmental Degradation of Materials in Nuclear Power Systems – Water Reactors. The purpose of this conference series is to foster an exchange of ideas about problems and their remedies in water-cooled nuclear power plants of today and the future. Contributions cover problems facing nickel-based alloys, stainless steels, pressure vessel and piping steels, zirconium alloys, and other alloys in water environments of relevance. Components covered include pressure boundary components, reactor vessels and internals, steam generators, fuel cladding, irradiated components, fuel storage containers, and balance of plant components and systems.
📒Energy Materials 2014 ✍ The Minerals, Metals & Materials Society (TMS)
✏Author : The Minerals, Metals & Materials Society (TMS)
✏Publisher : Springer
✏Release Date : 2017-03-16
✏Pages : 932
✏ISBN : 9783319487656
✏Available Language : English, Spanish, And French
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✏Energy Materials 2014 Book Summary :
✏Author :
✏Publisher :
✏Release Date : 2013
✏Pages :
✏ISBN : OCLC:967919948
✏Available Language : English, Spanish, And French
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✏SiC CMC Zircaloy 4 Nuclear Fuel Cladding Performance During 4 Point Tubular Bend Testing Book Summary : The U.S. Department of Energy Office of Nuclear Energy (DOE NE) established the Light Water Reactor Sustainability (LWRS) program to develop technologies and other solutions to improve the reliability, sustain the safety, and extend the life of current reactors. The Advanced LWR Nuclear Fuel Development Pathway in the LWRS program encompasses strategic research focused on improving reactor core economics and safety margins through the development of an advanced fuel cladding system. Recent investigations of potential options for "accident tolerant" nuclear fuel systems point to the potential benefits of silicon carbide (SiC) cladding. One of the proposed SiC-based fuel cladding designs being investigated incorporates a SiC ceramic matrix composite (CMC) as a structural material supplementing an internal Zircaloy-4 (Zr-4) liner tube, referred to as the hybrid clad design. Characterization of the advanced cladding designs will include a number of out-of-pile (nonnuclear) tests, followed by in-pile irradiation testing of the most promising designs. One of the out-of-pile characterization tests provides measurement of the mechanical properties of the cladding tube using four point bend testing. Although the material properties of the different subsystems (materials) will be determined separately, in this paper we present results of 4-point bending tests performed on fully assembled hybrid cladding tube mock-ups, an assembled Zr-4 cladding tube mock-up as a standard and initial testing results on bare SiC-CMC sleeves to assist in defining design parameters. The hybrid mock-up samples incorporated SiC-CMC sleeves fabricated with 7 polymer impregnation and pyrolysis (PIP) cycles. To provide comparative information; both 1- and 2-ply braided SiC-CMC sleeves were used in this development study. Preliminary stress simulations were performed using the BISON nuclear fuel performance code to show the stress distribution differences for varying lengths between loading points and clad configurations. The 2-ply sleeve samples show a higher bend momentum compared to those of the 1-ply sleeve samples. This is applicable to both the hybrid mock-up and bare SiC-CMC sleeve samples. Comparatively both the 1- and 2-ply hybrid mock-up samples showed a higher bend stiffness and strength compared with the standard Zr-4 mock-up sample. The characterization of the hybrid mock-up samples showed signs of distress and preliminary signs of fraying at the protective Zr-4 sleeve areas for the 1-ply SiC-CMC sleeve. In addition, the microstructure of the SiC matrix near the cracks at the region of highest compressive bending strain shows significant cracking and flaking. The 2-ply SiC-CMC sleeve samples showed a more bonded, cohesive SiC matrix structure. This cracking and fraying causes concern for increased fretting during the actual use of the design. Tomography was proven as a successful tool to identify open porosity during pre-test characterization. Although there is currently insufficient data to make conclusive statements regarding the overall merit of the hybrid cladding design, preliminary characterization of this novel design has been demonstrated.
📒Ceramic Materials For Energy Applications Iii ✍ Hua-Tay Lin
✏Author : Hua-Tay Lin
✏Publisher : John Wiley & Sons
✏Release Date : 2013-12-02
✏Pages : 176
✏ISBN : 9781118807859
✏Available Language : English, Spanish, And French
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✏Ceramic Materials for Energy Applications III Book Summary : Ceramic Engineering and Science Proceedings Volume 34, Issue 9 - Ceramic Materials for Energy Applications III A collection of 15 papers from The American Ceramic Society’s 37th International Conference on Advanced Ceramics and Composites, held in Daytona Beach, Florida, January 27-February 1, 2013. This issue includes papers presented in Symposia 6 - Advanced Materials and Technologies for Rechargeable Energy Storage; Symposium 13 - Advanced Ceramics and Composites for Sustainable Nuclear Energy and Fusion Energy; Focused Session 4 - Advanced Processing for Photonics and Energy; and the Engineering Summit of the Americas session.
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✏Publisher :
✏Release Date : 1993
✏Pages : 12
✏ISBN : PSU:000023716155
✏Available Language : English, Spanish, And French
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✏American National Standard Design Requirements for Light Water Reactor Fuel Handling Systems Book Summary :
✏Author :
✏Publisher :
✏Release Date : 2014
✏Pages :
✏ISBN : OCLC:1065862942
✏Available Language : English, Spanish, And French
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✏Systematic Technology Evaluation Program for SiC SiC Composite based Accident Tolerant LWR Fuel Cladding and Core Structures M2FT 14OR0202244 Book Summary : Fuels and core structures in the current light water reactors (LWR's) are vulnerable to catastrophic consequences in the event of loss of coolant or active cooling, as unfortunately evidenced by the March 2011 Fukushima Dai-ichi Nuclear Power Plant Accident [1-3]. This vulnerability is attributed primarily to the rapid oxidation kinetics of zirconium alloys in a water vapor environment at very high temperatures [1, 4]. Current LWR's use Zr alloys nearly exclusively as the materials for fuel cladding and core structures. Among the candidate alternative materials for the LWR fuel clads and core structures to enable so-called accident-tolerant fuels (ATF) and accident-tolerant cores (ATC), silicon carbide (SiC) - based materials, in particular continuous SiC fiber-reinforced SiC matrix ceramic composites (SiC/SiC composites or SiC composites), are considered to provide outstanding passive safety features in beyond-design basis severe accident scenarios [3, 5, 6]. The SiC/SiC composites are anticipated to provide additional benefits over the zirconium alloys, including the smaller neutron cross sections, general chemical inertness, ability to withstand higher fuel burn-ups and higher temperatures, exceptional inherent radiation resistance, lack of progressive irradiation growth, and low induced-activation / low decay heat [7]. SiC/SiC composites are finding specialty applications as industrial materials as they mature and their application technologies grow [8]. Moreover, SiC and SiC/SiC composites are among the materials that have most extensively been studied for the effects of irradiation for nuclear applications.
📒Tms 2018 147th Annual Meeting Exhibition Supplemental Proceedings ✍ The Minerals, Metals & Materials Society
✏Author : The Minerals, Metals & Materials Society
✏Publisher : Springer
✏Release Date : 2018-02-03
✏Pages : 961
✏ISBN : 9783319725260
✏Available Language : English, Spanish, And French
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✏TMS 2018 147th Annual Meeting Exhibition Supplemental Proceedings Book Summary : This collection features papers presented at the 147th Annual Meeting & Exhibition of The Minerals, Metals & Materials Society.
📒Regulatory And Technical Reports ✍ U.S. Nuclear Regulatory Commission. Policy and Publications Management Branch
✏Author : U.S. Nuclear Regulatory Commission. Policy and Publications Management Branch
✏Publisher :
✏Release Date : 1980
✏Pages :
✏ISBN : IND:30000089265262
✏Available Language : English, Spanish, And French
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✏Regulatory and Technical Reports Book Summary :
✏Author :
✏Publisher :
✏Release Date : 1980
✏Pages :
✏ISBN : MINN:30000010506370
✏Available Language : English, Spanish, And French
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✏Energy Research Abstracts Book Summary :
📒American National Standard For Light Water Reactors Fuel Assembly Mechanical Design And Evaluation ✍ American National Standards Institute
✏Author : American National Standards Institute
✏Publisher :
✏Release Date : 1996
✏Pages : 17
✏ISBN : PSU:000032733495
✏Available Language : English, Spanish, And French
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✏American National Standard for Light Water Reactors Fuel Assembly Mechanical Design and Evaluation Book Summary :
✏Author :
✏Publisher :
✏Release Date : 2016
✏Pages : 46
✏ISBN : OCLC:962177256
✏Available Language : English, Spanish, And French
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✏BISON Fuel Performance Analysis of FeCrAl Cladding with Updated Properties Book Summary : In order to improve the accident tolerance of light water reactor (LWR) fuel, alternative cladding materials have been proposed to replace zirconium (Zr)-based alloys. Of these materials, there is a particular focus on iron-chromium-aluminum (FeCrAl) alloys due to much slower oxidation kinetics in high-temperature steam than Zr-alloys. This should decrease the energy release due to oxidation and allow the cladding to remain integral longer in the presence of high temperature steam, making accident mitigation more likely. As a continuation of the development for these alloys, suitability for normal operation must also be demonstrated. This research is focused on modeling the integral thermo-mechanical performance of FeCrAl cladded fuel during normal reactor operation. Preliminary analysis has been performed to assess FeCrAl alloys (namely Alkrothal 720 and APMT) as a suitable fuel cladding replacement for Zr-alloys, using the MOOSE-based, finite-element fuel performance code BISON and the best available thermal-mechanical and irradiation-induced constitutive properties. These simulations identify the effects of the mechanical-stress and irradiation response of FeCrAl and provide a comparison with Zr-alloys. In comparing these clad materials, fuel rods have been simulated for normal reactor operation and simple steady-state operation. Normal reactor operating conditions target the cladding performance over the rod lifetime (~4 cycles) for the highest-power rod in the highest-power fuel assembly under reactor power maneuvering. The power histories and axial temperature profiles input into BISON were generated from a neutronics study on full-core reactivity equivalence for FeCrAl using the 3D full core simulator NESTLE. Evolution of the FeCrAl cladding behavior over time is evaluated by using steady-state operating conditions such as a simple axial power profile, a constant cladding surface temperature, and a constant fuel power history. The fuel rod designs and operating conditions used are based off the Peach Bottom BWR and design consideration was given to minimize the neutronic penalty of the FeCrAl cladding by changing fuel enrichment and cladding thickness. As this study progressed, systematic parametric analysis of the fuel and cladding creep responses were also performed.
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✏Publisher :
✏Release Date : 2014
✏Pages :
✏ISBN : OCLC:1066485820
✏Available Language : English, Spanish, And French
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✏STATUS OF HIGH FLUX ISOTOPE REACTOR IRRADIATION OF SILICON CARBIDE SILICON CARBIDE JOINTS Book Summary : 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.
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✏Release Date : 1976
✏Pages :
✏ISBN : STANFORD:36105061890575
✏Available Language : English, Spanish, And French
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✏Monthly Catalog of United States Government Publications Book Summary :
📒Transactions Of The American Nuclear Society ✍ American Nuclear Society
✏Author : American Nuclear Society
✏Publisher :
✏Release Date : 2000
✏Pages :
✏ISBN : UOM:39015047827905
✏Available Language : English, Spanish, And French
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✏Transactions of the American Nuclear Society Book Summary : Each volume contains proceedings of the annual conference of the American Nuclear Society.
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✏Release Date : 1987
✏Pages :
✏ISBN : CORNELL:31924083096739
✏Available Language : English, Spanish, And French
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✏Metals Abstracts Book Summary :
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✏Release Date : 1986
✏Pages :
✏ISBN : STANFORD:36105021589549
✏Available Language : English, Spanish, And French
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✏INIS Atomindex Book Summary :
✏Author :
✏Publisher :
✏Release Date : 1991
✏Pages :
✏ISBN : UOM:39015022357266
✏Available Language : English, Spanish, And French
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✏Engineered Materials Abstracts Book Summary :
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✏Release Date : 1981
✏Pages :
✏ISBN : STANFORD:36105014640879
✏Available Language : English, Spanish, And French
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✏Safety Science Abstracts Journal Book Summary :
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✏Release Date : 1986
✏Pages :
✏ISBN : MINN:30000008704458
✏Available Language : English, Spanish, And French
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✏Government Reports Annual Index Book Summary :
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✏Release Date : 1995
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✏ISBN : UOM:39015034545676
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✏The New Encyclopaedia Britannica Macropaedia Book Summary :
📒The New Encyclopaedia Britannica ✍ Encyclopaedia Britannica, inc
✏Author : Encyclopaedia Britannica, inc
✏Publisher :
✏Release Date : 1998
✏Pages :
✏ISBN : STANFORD:36105021693267
✏Available Language : English, Spanish, And French
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✏The New Encyclopaedia Britannica Book Summary :