Read Books Online and Download eBooks, EPub, PDF, Mobi, Kindle, Text Full Free.
Nitride Fuel Development At The Inl
Download Nitride Fuel Development At The Inl full books in PDF, epub, and Kindle. Read online Nitride Fuel Development At The Inl ebook anywhere anytime directly on your device. Fast Download speed and no annoying ads. We cannot guarantee that every ebooks is available!
Book Synopsis Nitride Fuel Development at the INL. by :
Download or read book Nitride Fuel Development at the INL. written by and published by . This book was released on 2007 with total page pages. Available in PDF, EPUB and Kindle. Book excerpt: A new method for fabricating nitride-based fuels for nuclear applications is under development at the Idaho National Laboratory (INL). A primary objective of this research is the development of a process that could be operated as an automated or semi-automated technique reducing costs, worker doses, and eventually improving the final product form. To achieve these goals the fabrication process utilizes a new cryo-forming technique to produce microspheres formed from sub-micron oxide powder to improve material handling issues, yield rapid kinetics for conversion to nitrides, and reduced material impurity levels within the nitride compounds. The microspheres are converted to a nitride form within a high temperature particle fluidizing bed using a carbothermic process that utilizes a hydrocarbon - hydrogen - nitrogen gas mixture. A new monitor and control system using differential pressure changes in the fluidizing gas allows for real-time monitoring and control of the spouted bed reactor during conversion. This monitor and control system can provide real-time data that is used to control the gas flow rates, temperatures, and gas composition to optimize the fluidization of the particle bed. The small size (0.5 μm) of the oxide powders in the microspheres dramatically increases the kinetics of the conversion process yielding reduced process times and temperatures. Initial studies using surrogate ZrO2 powder have yielded conversion efficiencies of 90 -95 % nitride formation with only small levels of oxide and carbide contaminants present. Further studies are being conducted to determine optimal gas mixture ratios, process time, and temperature range for providing complete conversion to a nitride form.
Book Synopsis Laboratory Directed Research and Development (LDRD) on Mono-uranium Nitride Fuel Development for SSTAR and Space Applications by : J. Ahn
Download or read book Laboratory Directed Research and Development (LDRD) on Mono-uranium Nitride Fuel Development for SSTAR and Space Applications written by J. Ahn and published by . This book was released on 2006 with total page 91 pages. Available in PDF, EPUB and Kindle. Book excerpt: The US National Energy Policy of 2001 advocated the development of advanced fuel and fuel cycle technologies that are cleaner, more efficient, less waste-intensive, and more proliferation resistant. The need for advanced fuel development is emphasized in on-going DOE-supported programs, e.g., Global Nuclear Energy Initiative (GNEI), Advanced Fuel Cycle Initiative (AFCI), and GEN-IV Technology Development. The Directorates of Energy & Environment (E & amp;E) and Chemistry & Material Sciences (C & amp;MS) at Lawrence Livermore National Laboratory (LLNL) are interested in advanced fuel research and manufacturing using its multi-disciplinary capability and facilities to support a design concept of a small, secure, transportable, and autonomous reactor (SSTAR). The E & E and C & MS Directorates co-sponsored this Laboratory Directed Research & Development (LDRD) Project on Mono-Uranium Nitride Fuel Development for SSTAR and Space Applications. In fact, three out of the six GEN-IV reactor concepts consider using the nitride-based fuel, as shown in Table 1. SSTAR is a liquid-metal cooled, fast reactor. It uses nitride fuel in a sealed reactor vessel that could be shipped to the user and returned to the supplier having never been opened in its long operating lifetime. This sealed reactor concept envisions no fuel refueling nor on-site storage of spent fuel, and as a result, can greatly enhance proliferation resistance. However, the requirement for a sealed, long-life core imposes great challenges to research and development of the nitride fuel and its cladding. Cladding is an important interface between the fuel and coolant and a barrier to prevent fission gas release during normal and accidental conditions. In fabricating the nitride fuel rods and assemblies, the cladding material should be selected based on its the coolant-side corrosion properties, the chemical/physical interaction with the nitride fuel, as well as their thermal and neutronic properties. The US NASA space reactor, the SP-100 was designed to use mono-uranium nitride fuel. Although the SP-100 reactor was not commissioned, tens of thousand of nitride fuel pellets were manufactured and lots of them, cladded in Nb-1-Zr had been irradiated in fast test reactors (FFTF and EBR-II) with good irradiation results. The Russian Naval submarines also use nitride fuel with stainless steel cladding (HT-9) in Pb-Bi coolant. Although the operating experience of the Russian submarine is not readily available, such combination of fuel, cladding and coolant has been proposed for a commercial-size liquid-metal cooled fast reactor (BREST-300). Uranium mono-nitride fuel is studied in this LDRD Project due to its favorable properties such as its high actinide density and high thermal conductivity. The thermal conductivity of mono-nitride is 10 times higher than that of oxide (23 W/m-K for UN vs. 2.3 W/m-K for UO{sub 2} at 1000 K) and its melting temperature is much higher than that of metal fuel (2630 C for UN vs. 1132 C for U metal). It also has relatively high actinide density, (13.51 gU/cm{sup 3} in UN vs. 9.66 gU/cm{sup 3} in UO{sub 2}) which is essential for a compact reactor core design. The objective of this LDRD Project is to: (1) Establish a manufacturing capability for uranium-based ceramic nuclear fuel, (2) Develop a computational capability to analyze nuclear fuel performance, (3) Develop a modified UN-based fuel that can support a compact long-life reactor core, and (4) Collaborate with the Nuclear Engineering Department of UC Berkeley on nitride fuel reprocessing and disposal in a geologic repository.
Book Synopsis Hauordnung für die Staatsforsten im Regierungsbezirk Kassel by :
Download or read book Hauordnung für die Staatsforsten im Regierungsbezirk Kassel written by and published by . This book was released on 1929 with total page 19 pages. Available in PDF, EPUB and Kindle. Book excerpt:
Book Synopsis Nitride Fuel Development Using Cryo-process Technique by :
Download or read book Nitride Fuel Development Using Cryo-process Technique written by and published by . This book was released on 2007 with total page pages. Available in PDF, EPUB and Kindle. Book excerpt: A new cryo-process technique has been developed for the fabrication of advanced fuel for nuclear systems. The process uses a new cryo-processing technique whereby small, porous microspheres (
Book Synopsis Summary of Recent Uranium Nitride Fuel Research and Development by :
Download or read book Summary of Recent Uranium Nitride Fuel Research and Development written by and published by . This book was released on 1962 with total page pages. Available in PDF, EPUB and Kindle. Book excerpt: Declassified 18 Sep 1973. Uranium nitride was developed as a high- temperature fuel for liquid metal-cooled reactors. The development included fabrication studies, environmental compatibility, thermal stability, hot hardness, and irradiation testing. Besides pure UN, mixtures of UN and ZrN were also investigated. (DLC).
Book Synopsis Performance Analysis of a Mixed Nitride Fuel System for an Advanced Liquid Metal Reactor by :
Download or read book Performance Analysis of a Mixed Nitride Fuel System for an Advanced Liquid Metal Reactor written by and published by . This book was released on 1990 with total page 8 pages. Available in PDF, EPUB and Kindle. Book excerpt: The conceptual development and analysis of a proposed mixed nitride driver and blanket fuel system for a prototypic advanced liquid metal reactor design has been performed. As a first step, an intensive literature survey was completed on the development and testing of nitride fuel systems. Based on the results of this survey, prototypic mixed nitride fuel and blanket pins were designed and analyzed using the SIEX computer code. The analysis predicted that the nitride fuel consistently operated at peak temperatures and cladding strain levels that compared quite favorably with competing fuel designs. These results, along with data available in the literature on nitride fuel performance, indicate that a nitride fuel system should offer enhanced capabilities for advanced liquid metal reactors. 13 refs., 10 figs., 2 tabs.
Book Synopsis Uranium Nitride Fuel Development, SNAP-50 by :
Download or read book Uranium Nitride Fuel Development, SNAP-50 written by and published by . This book was released on 1965 with total page pages. Available in PDF, EPUB and Kindle. Book excerpt: Declassified 12 Sep 1973. Development of UN in four major areas is presented: 1) powder synthesis and powder metallurgical fabrication, 2) physical and mechanical property testing, 3) fuel-cladding compatibility and barrier development, and 4) irradiation evaluation. Synthesis and sintering methods for producing large quantities of high-densily and high-purity UN are described. Results of experimental studies of the following properties are summarized: melting point, thermal conductivity, thermal expansion, vapor pressure, and thermodynamics, heat capacity, hot hardness, compressive creep, nitrogen self- diffusion, and electrical resistivity. Compatibility tests are described which demonstrated the need for a diffusion barrier. Lithium soak tests for up to 11,000 h at 2200 deg F demonstrated the stability and practicality of vapor- deposited tungsten-lined Nb--1 Zr alloy over the projected life and temperature of SNAP50. Similar static tests of purposely defected simulated fuel pins indicate a relatively high degree of stability of UN towards lithium. Instrumented capsule irradiation tests of simulated Nb--1 Zr alloy - clad fuel pins are described under 2 Mw(t) and 8 Mw(t) SNAP-50 conditions. Under 8 Mw(t) conditions, 20% fission gas release and 2% diametral cladding growth were observed after 2750 h at 2200 deg F (2.0 at.% U burnup). In-pile operation under 2 Mw(t) conditions was achieved for 5940 h at 2000 deg F (1.0 at.% U burnup) and 3360 h at 2200 deg F (1.5 at.% U burnup) while experiencing less than 0.2% fission gas release and less than 0.4% diametral growth. (66 figures) (auth).
Author :United States. Congress. House. Committee on Appropriations. Subcommittee on Energy and Water Development Publisher : ISBN 13 : Total Pages :1840 pages Book Rating :4.F/5 ( download)
Book Synopsis Energy and Water Development Appropriations for 2009 by : United States. Congress. House. Committee on Appropriations. Subcommittee on Energy and Water Development
Download or read book Energy and Water Development Appropriations for 2009 written by United States. Congress. House. Committee on Appropriations. Subcommittee on Energy and Water Development and published by . This book was released on 2008 with total page 1840 pages. Available in PDF, EPUB and Kindle. Book excerpt:
Book Synopsis Nitride Fuel Modeling Recommendation for Nitride Fuel Material Property Measurement Priority by : Richard Moore
Download or read book Nitride Fuel Modeling Recommendation for Nitride Fuel Material Property Measurement Priority written by Richard Moore and published by . This book was released on 2005 with total page pages. Available in PDF, EPUB and Kindle. Book excerpt: The purpose of this effort was to provide the basis for a model that effectively predicts nitride fuel behavior. Material property models developed for the uranium nitride fuel system have been used to approximate the general behavior of nitride fuels with specific property models for the transuranic nitride fuels utilized as they become available. The AFCI fuel development program now has the means for predicting the behavior of the transuranic nitride fuel compositions. The key data and models needed for input into this model include:Thermal conductivity with burnupFuel expansion coefficientFuel swelling with burnupFission gas release with burnup. Although the fuel performance model is a fully functional FEA analysis tool, it is limited by the input data and models.
Book Synopsis Reactor Fuels, Materials and Systems under Extreme Environments by : Wenzhong Zhou
Download or read book Reactor Fuels, Materials and Systems under Extreme Environments written by Wenzhong Zhou and published by Frontiers Media SA. This book was released on 2022-03-25 with total page 360 pages. Available in PDF, EPUB and Kindle. Book excerpt:
Book Synopsis Comprehensive Nuclear Materials by :
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
Book Synopsis Medical Isotope Production Without Highly Enriched Uranium by : National Research Council
Download or read book Medical Isotope Production Without Highly Enriched Uranium written by National Research Council and published by National Academies Press. This book was released on 2009-06-27 with total page 220 pages. Available in PDF, EPUB and Kindle. Book excerpt: This book is the product of a congressionally mandated study to examine the feasibility of eliminating the use of highly enriched uranium (HEU2) in reactor fuel, reactor targets, and medical isotope production facilities. The book focuses primarily on the use of HEU for the production of the medical isotope molybdenum-99 (Mo-99), whose decay product, technetium-99m3 (Tc-99m), is used in the majority of medical diagnostic imaging procedures in the United States, and secondarily on the use of HEU for research and test reactor fuel. The supply of Mo-99 in the U.S. is likely to be unreliable until newer production sources come online. The reliability of the current supply system is an important medical isotope concern; this book concludes that achieving a cost difference of less than 10 percent in facilities that will need to convert from HEU- to LEU-based Mo-99 production is much less important than is reliability of supply.
Book Synopsis Viability of Inert Matrix Fuel in Reducing Plutonium Amounts in Reactors by : International Atomic Energy Agency
Download or read book Viability of Inert Matrix Fuel in Reducing Plutonium Amounts in Reactors written by International Atomic Energy Agency and published by IAEA. This book was released on 2006 with total page 100 pages. Available in PDF, EPUB and Kindle. Book excerpt: The reactors around the world have produced more than 2000 tonnes of plutonium, contained in spent fuel or as separated forms through reprocessing. Disposition of fissile materials has become a primary concern of nuclear non-proliferation efforts worldwide. There is a significant interest in IAEA Member States to develop proliferation resistant nuclear fuel cycles for incineration of plutonium such as inert matrix fuels (IMFs). This publication reviews the status of potential IMF candidates and describes several identified candidate materials for both fast and thermal reactors: MgO, ZrO2, SiC, Zr alloy, SiAl, ZrN; some of these have undergone test irradiations and post irradiation examination. Also discussed are modelling of IMF fuel performance and safety analysis. System studies have identified strategies for both implementation of IMF fuel as homogeneous or heterogeneous phases, as assemblies or core loadings and in existing reactors in the shorter term, as well as in new reactors in the longer term.
Book Synopsis Application of Self-Propagating High Temperature Synthesis to the Fabrication of Actinide Bearing Nitride and Other Ceramic Nuclear Fuels by :
Download or read book Application of Self-Propagating High Temperature Synthesis to the Fabrication of Actinide Bearing Nitride and Other Ceramic Nuclear Fuels written by and published by . This book was released on 2009 with total page 5 pages. Available in PDF, EPUB and Kindle. Book excerpt: The project uses an exothermic combustion synthesis reaction, termed self-propagating high-temperature synthesis (SHS), to produce high quality, reproducible nitride fuels and other ceramic type nuclear fuels (cercers and cermets, etc.) in conjunction with the fabrication of transmutation fuels. The major research objective of the project is determining the fundamental SHS processing parameters by first using manganese as a surrogate for americium to produce dense Zr-Mn-N ceramic compounds. These fundamental principles will then be transferred to the production of dense Zr-Am-N ceramic materials. A further research objective in the research program is generating fundamental SHS processing data to the synthesis of (i) Pu-Am-Zr-N and (ii) U-Pu-Am-N ceramic fuels. In this case, Ce will be used as the surrogate for Pu, Mn as the surrogate for Am, and depleted uranium as the surrogate for U. Once sufficient fundamental data has been determined for these surrogate systems, the information will be transferred to Idaho National Laboratory (INL) for synthesis of Zr-Am-N, Pu-Am-Zr-N and U-Pu-Am-N ceramic fuels. The high vapor pressures of americium (Am) and americium nitride (AmN) are cause for concern in producing nitride ceramic nuclear fuel that contains Am. Along with the problem of Am retention during the sintering phases of current processing methods, are additional concerns of producing a consistent product of desirable homogeneity, density and porosity. Similar difficulties have been experienced during the laboratory scale process development stage of producing metal alloys containing Am wherein compact powder sintering methods had to be abandoned. Therefore, there is an urgent need to develop a low-temperature or low-heat fuel fabrication process for the synthesis of Am-containing ceramic fuels. Self-propagating high temperature synthesis (SHS), also called combustion synthesis, offers such an alternative process for the synthesis of Am nitride fuels. Although SHS takes thermodynamic advantage of the high combustion temperatures of these exothermic SHS reactions to synthesize the required compounds, the very fast heating, reaction and cooling rates can kinetically generate extremely fast reaction rates and facilitate the retention of volatile species within the rapidly propagating SHS reaction front. The initial objective of the research program is to use Mn as the surrogate for Am to synthesize a reproducible, dense, high quality Zr-Mn-N ceramic compound. Having determined the fundamental SHS reaction parameters and optimized SHS processing steps using Mn as the surrogate for Am, the technology will be transferred to Idaho National Laboratory to successfully synthesize a high quality Zr-Am-N ceramic fuel.
Book Synopsis Internationalization of the Nuclear Fuel Cycle by : Russian Academy of Sciences
Download or read book Internationalization of the Nuclear Fuel Cycle written by Russian Academy of Sciences and published by National Academies Press. This book was released on 2009-01-26 with total page 172 pages. Available in PDF, EPUB and Kindle. Book excerpt: The so-called nuclear renaissance has increased worldwide interest in nuclear power. This potential growth also has increased, in some quarters, concern that nonproliferation considerations are not being given sufficient attention. In particular, since introduction of many new power reactors will lead to requiring increased uranium enrichment services to provide the reactor fuel, the proliferation risk of adding enrichment facilities in countries that do not have them now led to proposals to provide the needed fuel without requiring indigenous enrichment facilities. Similar concerns exist for reprocessing facilities. Internationalization of the Nuclear Fuel Cycle summarizes key issues and analyses of the topic, offers some criteria for evaluating options, and makes findings and recommendations to help the United States, the Russian Federation, and the international community reduce proliferation and other risks, as nuclear power is used more widely. This book is intended for all those who are concerned about the need for assuring fuel for new reactors and at the same time limiting the spread of nuclear weapons. This audience includes the United States and Russia, other nations that currently supply nuclear material and technology, many other countries contemplating starting or growing nuclear power programs, and the international organizations that support the safe, secure functioning of the international nuclear fuel cycle, most prominently the International Atomic Energy Agency.
Author :NEA Nuclear Science Committee. Task Force on Scientific Issues Related to Fuel Behaviour Publisher :OECD ISBN 13 : Total Pages :96 pages Book Rating :4.3/5 (91 download)
Book Synopsis Scientific Issues in Fuel Behaviour by : NEA Nuclear Science Committee. Task Force on Scientific Issues Related to Fuel Behaviour
Download or read book Scientific Issues in Fuel Behaviour written by NEA Nuclear Science Committee. Task Force on Scientific Issues Related to Fuel Behaviour and published by OECD. This book was released on 1995 with total page 96 pages. Available in PDF, EPUB and Kindle. Book excerpt: Dated January 1995
Book Synopsis A Modified Nitride-Based Fuel for Long Core Life and Proliferation Resistance by : J. Choi
Download or read book A Modified Nitride-Based Fuel for Long Core Life and Proliferation Resistance written by J. Choi and published by . This book was released on 2003 with total page 8 pages. Available in PDF, EPUB and Kindle. Book excerpt: A modified nitride-based uranium fuel to support the small, secured, transportable, and autonomous reactor (SSTAR) concept is initiated at Lawrence Livermore National laboratory (LLNL). This project centers on the evaluation of modified uranium nitride fuels imbedded with other inert (e.g. ZrN), neutron-absorbing (e.g. HfN), or breeding (e.g. ThN) nitrides to enhance the fuel properties to achieve long core life with a compact reactor design. A long-life fuel could minimize the need for on-site refueling and spent-fuel storage. As a result, it could significantly improve the proliferation resistance of the reactor/fuel systems. This paper discusses the potential benefits and detriments of modified nitride-based fuels using the criteria of compactness, long-life, proliferation resistance, fuel safety, and waste management. Benefits and detriments are then considered in recommending a select set of compositions for further study.
