Download or read book Systematic Technology Evaluation Program for SiC SiC Composite based Accident Tolerant LWR Fuel Cladding and Core Structures written by and published by . This book was released on 2014 with total page pages. Available in PDF, EPUB and Kindle. Book excerpt:

Download or read book Systematic Technology Evaluation Program for SiC SiC Composite based Accident Tolerant LWR Fuel Cladding and Core Structures M2FT 14OR0202244 written by and published by . This book was released on 2014 with total page pages. Available in PDF, EPUB and Kindle. Book excerpt: 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.

Download or read book Systematic Technology Evaluation Program for SiC SiC Composite based Accident tolerant LWR Fuel Cladding and Core Structures written by and published by . This book was released on 2015 with total page 51 pages. Available in PDF, EPUB and Kindle. Book excerpt: Fuels and core structures in current light water reactors (LWR's) are vulnerable to catastrophic failure in severe accidents as unfortunately evidenced by the March 2011 Fukushima Dai-ichi Nuclear Power Plant Accident. This vulnerability is attributed primarily to the rapid oxidation kinetics of zirconium alloys in a water vapor environment at very high temperatures. Zr alloys are the primary material in LWR cores except for the fuel itself. Therefore, alternative materials with reduced oxidation kinetics as compared to zirconium alloys are sought to enable enhanced accident-tolerant fuels and cores.

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 Report on Status of Execution of SiC Step Document written by and published by . This book was released on 2015 with total page 18 pages. Available in PDF, EPUB and Kindle. Book excerpt: Advanced fuel claddings made entirely or mainly of silicon carbide (SiC) ceramics and/or composites are considered very attractive elements of the accident-tolerant fuels for the light water reactors. In order to translate the promise of SiC composite materials into a reliable fuel cladding, a coordinated program of component level design and materials development must be carried out with many key feasibility issues addressed a-priori to inform the process. With the primary objective of developing a draft blueprint of a technical program that addresses the critical feasibility issues; assesses design and performance issues related with manufacturing, operating, and off-normal events; and advances the technological readiness levels in essential technology elements, a draft plan for the Systematic Technology Evaluation Program for SiC/SiC Composite Accident-Tolerant LWR Fuel Cladding and Core Structures was developed in the FY-14 Advanced Fuels Campaign of the U.S. Department of Energy's Fuel Cycles Research and Development Program. This document summarizes the status of execution of the technical plan within the activities at the Oak Ridge National Laboratory.

Download or read book Ceramic Materials for Energy Applications V written by Josef Matyas and published by John Wiley & Sons. This book was released on 2015-12-23 with total page 164 pages. Available in PDF, EPUB and Kindle. Book excerpt: The Ceramic Engineering and Science Proceeding has been published by The American Ceramic Society since 1980. This series contains a collection of papers dealing with issues in both traditional ceramics (i.e., glass, whitewares, refractories, and porcelain enamel) and advanced ceramics. Topics covered in the area of advanced ceramic include bioceramics, nanomaterials, composites, solid oxide fuel cells, mechanical properties and structural design, advanced ceramic coatings, ceramic armor, porous ceramics, and more.

Download or read book Evaluation of Multilayer Silicon Carbide Composite Cladding Under Loss of Coolant Accident Conditions written by Gregory Welch Daines and published by . This book was released on 2016 with total page 164 pages. Available in PDF, EPUB and Kindle. Book excerpt: Silicon carbide (SiC) has been proposed as an alternative to zirconium alloys used in current light water reactor (LWR) fuel cladding because it exhibits superior corrosion characteristics, high-temperature strength, and a 1000°C higher melting temperature, all of which are important during a loss of coolant accident (LOCA). To improve the performance of SiC cladding, a multilayered architecture consisting of layers of monolithic SiC (mSiC) and SiC/SiC ceramic matrix composite (CMC) has been proposed. In this work, the mechanical performance of both the tubing and the endplug joint of two-layer SiC cladding is investigated under conditions associated with the LOCA. Specifically, SiC cladding mechanical performance is investigated after exposure to 1,400°C steam and after quenching from 1,200°C into either 100°C or 90°C atmospheric-pressure water. The samples consist of two-layer SiC, with an inner SiC/SiC CMC layer and an outer monolith SiC layer. The relationship between mechanical performance and sample architecture is investigated through ceramography and internal void characterization. The two-layered SiC cladding design offered an as-received failure hoop stress of about 600 MPa, with little strength reduction due to thermal shock, and the tube failure hoop stress remained above 200 MPa after 48 hour high-temperature steam oxidation. The cladding showed pseudo-ductile behavior and failed in a non-frangible manner. The designs investigated for joint strength offered as-received burst strength above 30 MPa, although the impact of thermal shock and oxidation showed possible dependence on architecture. Overall, the cladding showed promising accident-tolerant performance. Because the implementation of SiC is complicated by the need for an open gap and low plenum pressure, thorium-based mixed oxides (MOX) are a promising fuel for SiC cladding because they have higher thermal conductivity and lower fission gas release (FGR). Previous efforts at MIT have modified the FRAPCON code to include thorium MOX fuel. In this work, the fission gas release and thermal conductivity models of FRAPCON-3.4-MIT are validated against published data. The results of this validation indicate a need to update the FGR model, which was accomplished in this work.

Download or read book SiC MODIFICATIONS TO MELCOR FOR SEVERE ACCIDENT ANALYSIS APPLICATIONS written by and published by . This book was released on 2013 with total page pages. Available in PDF, EPUB and Kindle. Book excerpt: The Department of Energy (DOE) Office of Nuclear Energy (NE) Light Water Reactor (LWR) Sustainability Program encompasses strategic research focused on improving reactor core economics and safety margins through the development of an advanced fuel cladding system. The Fuels Pathway within this program focuses on fuel system components outside of the fuel pellet, allowing for alteration of the existing zirconium-based clad system through coatings, addition of ceramic sleeves, or complete replacement (e.g. fully ceramic cladding). The DOE-NE Fuel Cycle Research & Development (FCRD) Advanced Fuels Campaign (AFC) is also conducting research on materials for advanced, accident tolerant fuels and cladding for application in operating LWRs. To aide in this assessment, a silicon carbide (SiC) version of the MELCOR code was developed by substituting SiC in place of Zircaloy in MELCOR's reactor core oxidation and material property routines. The purpose of this development effort is to provide a numerical capability for estimating the safety advantages of replacing Zr-alloy components in LWRs with SiC components. This modified version of the MELCOR code was applied to the Three Mile Island (TMI-2) plant accident. While the results are considered preliminary, SiC cladding showed a dramatic safety advantage over Zircaloy cladding during this accident.

Download or read book Early Implementation of SiC Cladding Fuel Performance Models in BISON written by and published by . This book was released on 2015 with total page 38 pages. Available in PDF, EPUB and Kindle. Book excerpt: SiC-based ceramic matrix composites (CMCs) [5-8] are being developed and evaluated internationally as potential LWR cladding options. These development activities include interests within both the DOE-NE LWR Sustainability (LWRS) Program and the DOE-NE Advanced Fuels Campaign. The LWRS Program considers SiC ceramic matrix composites (CMCs) as offering potentially revolutionary gains as a cladding material, with possible benefits including more efficient normal operating conditions and higher safety margins under accident conditions [9]. Within the Advanced Fuels Campaign, SiC-based composites are a candidate ATF cladding material that could achieve several goals, such as reducing the rates of heat and hydrogen generation due to lower cladding oxidation rates in HT steam [10]. This work focuses on the application of SiC cladding as an ATF cladding material in PWRs, but these work efforts also support the general development and assessment of SiC as an LWR cladding material in a much broader sense.

Download or read book Design Optimization of Advanced PWR SiC SiC Fuel Cladding for Enhanced Tolerance of Loss of Coolant Conditions written by Pierre Guenoun (S.M.) and published by . This book was released on 2016 with total page 68 pages. Available in PDF, EPUB and Kindle. Book excerpt: Limited data has been published (especially on experimental work) on integrated multilayer SiC/SiC prototypical fuel cladding. In this work the mechanical performance of three unique architectures of three-layer silicon carbide (SiC) composite cladding is experimentally investigated under conditions associated with the loss of coolant accident (LOCA), and analytically under various conditions. Specifically, this work investigates SiC cladding mechanical performance after exposure to 1,400°C steam for 48 hours and after thermal shock induced by quenching from 1,200°C into either 100°C or 90°C water. Mechanical performance characteristics are thereafter correlated with sample architecture through void characterization and ceramography. The series with a reduced thickness did not have a pseudo-ductile regime due to overloading of the composite layer. The presence of the axial tow did not yield significant difference in the mechanical behavior most likely because samples were tested in the hoop direction. While as-received and quenched samples behaved similarly (pseudo ductile failure except for one series), non-frangible brittle failure (single-crack failure with no release of debris) was systematically observed after oxidation due to silica buildup in the inner voids of the ceramic matrix composite (CMC) layer. Overall, thermal shock had limited influence on sample mechanical characteristics and oxidation resulted in the formation of silica on the inner wall of the CMC voids leading to the weakening of the monolith matrix and brittle fracture. Stress field in the cladding design is simulated by finite element analysis under service and shutdown conditions at both the core's middle height and at the end of the fuel rod. Stresses in the fuel region are driven by the thermal gradient that creates stresses predominantly from irradiation induced swelling. At the endplug, constraints are mainly mechanical. Stress calculations show high sensitivity to the data scatter and especially swelling and thermal conductivity. No cladding with the design studied here can survive either service or shutdown conditions because of the high irradiation-induced tensile stresses that develop in the hot inner monolith layer. It is shown that this peak tensile stress can be alleviated by adjusting the swelling level of the different layers. The addition of an under-swelling material such as PyC or Si can reduce the monolith tensile stress by 10%. With a composite that swells 10% less than the monolith, the stress is reduced by 20%.

Download or read book Assessment of Silicon Carbide Composites for Advanced Salt Cooled Reactors written by and published by . This book was released on 2007 with total page pages. Available in PDF, EPUB and Kindle. Book excerpt: The Advanced High-Temperature Reactor (AHTR) is a new reactor concept that uses a liquid fluoride salt coolant and a solid high-temperature fuel. Several alternative fuel types are being considered for this reactor. One set of fuel options is the use of pin-type fuel assemblies with silicon carbide (SiC) cladding. This report provides (1) an initial viability assessment of using SiC as fuel cladding and other in-core components of the AHTR, (2) the current status of SiC technology, and (3) recommendations on the path forward. Based on the analysis of requirements, continuous SiC fiber-reinforced, chemically vapor-infiltrated SiC matrix (CVI SiC/SiC) composites are recommended as the primary option for further study on AHTR fuel cladding among various industrially available forms of SiC. Critical feasibility issues for the SiC-based AHTR fuel cladding are identified to be (1) corrosion of SiC in the candidate liquid salts, (2) high dose neutron radiation effects, (3) static fatigue failure of SiC/SiC, (4) long-term radiation effects including irradiation creep and radiation-enhanced static fatigue, and (5) fabrication technology of hermetic wall and sealing end caps. Considering the results of the issues analysis and the prospects of ongoing SiC research and development in other nuclear programs, recommendations on the path forward is provided in the order or priority as: (1) thermodynamic analysis and experimental examination of SiC corrosion in the candidate liquid salts, (2) assessment of long-term mechanical integrity issues using prototypical component sections, and (3) assessment of high dose radiation effects relevant to the anticipated operating condition.

Download or read book Technique Development for Modulus Microcracking Hermeticity and Coating Evaluation Capability Characterization of SiC SiC Tubes written by and published by . This book was released on 2016 with total page 36 pages. Available in PDF, EPUB and Kindle. Book excerpt: Driven by the need to enlarge the safety margins of nuclear fission reactors in accident scenarios, research and development of accident-tolerant fuel has become an important topic in the nuclear engineering and materials community. A continuous-fiber SiC/SiC composite is under consideration as a replacement for traditional zirconium alloy cladding owing to its high-temperature stability, chemical inertness, and exceptional irradiation resistance. An important task is the development of characterization techniques for SiC/SiC cladding, since traditional work using rectangular bars or disks cannot directly provide useful information on the properties of SiC/SiC composite tubes for fuel cladding applications. At Oak Ridge National Laboratory, experimental capabilities are under development to characterize the modulus, microcracking, and hermeticity of as-fabricated, as-irradiated SiC/SiC composite tubes. Resonant ultrasound spectroscopy has been validated as a promising technique to evaluate the elastic properties of SiC/SiC composite tubes and microcracking within the material. A similar technique, impulse excitation, is efficient in determining the basic mechanical properties of SiC bars prepared by chemical vapor deposition; it also has potential for application in studying the mechanical properties of SiC/SiC composite tubes. Complete evaluation of the quality of the developed coatings, a major mitigation strategy against gas permeation and hydrothermal corrosion, requires the deployment of various experimental techniques, such as scratch indentation, tensile pulling-off tests, and scanning electron microscopy. In addition, a comprehensive permeation test station is being established to assess the hermeticity of SiC/SiC composite tubes and to determine the H/D/He permeability of SiC/SiC composites. This report summarizes the current status of the development of these experimental capabilities.

Download or read book Advances in SiC SiC Ceramic Composites written by Akira Kohyama and published by John Wiley & Sons. This book was released on 2012-04-11 with total page 400 pages. Available in PDF, EPUB and Kindle. Book excerpt: This transactions volume contains 33 papers from the CREST International Symposium on SiC/SiC Composite Materials Research and Development and Its Application to Advanced Energy Systems held May 20-22, 2002 in Kyoto, Japan. Chapters include Processing for SiC/SiC Composites; Processing for SiC/SiC Composite Constituent; Characterization of Thermomechanical Performance; and Joining Technologies for Advanced Energy Applications. 373 pages.

Download or read book Nuclear Fuel Safety Criteria written by OECD Nuclear Energy Agency and published by OECD Publishing. This book was released on 2001 with total page 74 pages. Available in PDF, EPUB and Kindle. Book excerpt: Presents brief descriptions of 20 fuel-related safety criteria along with both the rationale for having such criteria and possible new design and operational issues which could have an effect on them.

Download or read book Final Report on Accident Tolerant Fuel Performance Analysis of APMT Steel Clad UO2 Fuel and APMT Steel Clad UN U3Si5 Fuel Concepts written by and published by . This book was released on 2014 with total page pages. Available in PDF, EPUB and Kindle. Book excerpt: In FY2014 our group completed and documented analysis of new Accident Tolerant Fuel (ATF) concepts using BISON. We have modeled the viability of moving from Zircaloy to stainless steel cladding in traditional light water reactors (LWRs). We have explored the reactivity penalty of this change using the MCNP-based burnup code Monteburns, while attempting to minimize this penalty by increasing the fuel pellet radius and decreasing the cladding thickness. Fuel performance simulations using BISON have also been performed to quantify changes to structural integrity resulting from thinner stainless steel claddings. We account for thermal and irradiation creep, fission gas swelling, thermal swelling and fuel relocation in the models for both Zircaloy and stainless steel claddings. Additional models that account for the lower oxidation stainless steel APMT are also invoked where available. Irradiation data for HT9 is used as a fallback in the absence of appropriate models. In this study the isotopic vectors within each natural element are varied to assess potential reactivity gains if advanced enrichment capabilities were levied towards cladding technologies. Recommendations on cladding thicknesses for a robust cladding as well as the constitutive components of a less penalizing composition are provided. In the first section (section 1-3), we present results accepted for publication in the 2014 TOPFUEL conference regarding the APMT/UO2 ATF concept (J. Galloway & C. Unal, Accident Tolerant and Neutronically Favorable LWR Cladding, Proceedings of WRFPM 2014, Sendai, Japan, Paper No. 1000050). Next we discuss our preliminary findings from the thermo-mechanical analysis of UN-U3Si5 fuel with APMT clad. In this analysis we used models developed from limited data that need to be updated when the irradiation data from ATF-1 test is available. Initial results indicate a swelling rate less than 1.5% is needed to prevent excessive clad stress.

Download or read book Accident Tolerant Fuel and Cladding Assessment written by and published by . This book was released on 2014 with total page pages. Available in PDF, EPUB and Kindle. Book excerpt:

Download or read book Structural Materials for Generation IV Nuclear Reactors written by Pascal Yvon and published by Woodhead Publishing. This book was released on 2016-08-27 with total page 686 pages. Available in PDF, EPUB and Kindle. Book excerpt: Operating at a high level of fuel efficiency, safety, proliferation-resistance, sustainability and cost, generation IV nuclear reactors promise enhanced features to an energy resource which is already seen as an outstanding source of reliable base load power. The performance and reliability of materials when subjected to the higher neutron doses and extremely corrosive higher temperature environments that will be found in generation IV nuclear reactors are essential areas of study, as key considerations for the successful development of generation IV reactors are suitable structural materials for both in-core and out-of-core applications. Structural Materials for Generation IV Nuclear Reactors explores the current state-of-the art in these areas. Part One reviews the materials, requirements and challenges in generation IV systems. Part Two presents the core materials with chapters on irradiation resistant austenitic steels, ODS/FM steels and refractory metals amongst others. Part Three looks at out-of-core materials. Structural Materials for Generation IV Nuclear Reactors is an essential reference text for professional scientists, engineers and postgraduate researchers involved in the development of generation IV nuclear reactors. - Introduces the higher neutron doses and extremely corrosive higher temperature environments that will be found in generation IV nuclear reactors and implications for structural materials - Contains chapters on the key core and out-of-core materials, from steels to advanced micro-laminates - Written by an expert in that particular area