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Author: Brian R. T. Frost Publisher: Elsevier ISBN: 1483155250 Category : Technology & Engineering Languages : en Pages : 284
Book Description
Nuclear Fuel Elements: Design, Fabrication and Performance is concerned with the design, fabrication, and performance of nuclear fuel elements, with emphasis on fast reactor fuel elements. Topics range from fuel types and the irradiation behavior of fuels to cladding and duct materials, fuel element design and modeling, fuel element performance testing and qualification, and the performance of water reactor fuels. Fast reactor fuel elements, research and test reactor fuel elements, and unconventional fuel elements are also covered. This volume consists of 12 chapters and begins with an overview of nuclear reactors and fuel elements, as well as fuel element design and development based on the reactor operator's approach, materials scientist's approach, and interdisciplinary approach. The reader is then introduced to different types of nuclear fuels and their irradiation behavior, considerations for using cladding and duct materials in fuel element design and development, and fuel element design and modeling. The chapters that follow focus on the testing of fuel element performance, experimental techniques and equipment for testing fuel element designs, and the performance of fuels for water reactors. Fuel elements for gas-cooled reactors, fast reactors, and research and test reactors are also described. The book concludes with an assessment of unconventional fuel elements. This book will be useful to fuel element technologists as well as materials scientists and engineers.
Author: Brian R. T. Frost Publisher: Elsevier ISBN: 1483155250 Category : Technology & Engineering Languages : en Pages : 284
Book Description
Nuclear Fuel Elements: Design, Fabrication and Performance is concerned with the design, fabrication, and performance of nuclear fuel elements, with emphasis on fast reactor fuel elements. Topics range from fuel types and the irradiation behavior of fuels to cladding and duct materials, fuel element design and modeling, fuel element performance testing and qualification, and the performance of water reactor fuels. Fast reactor fuel elements, research and test reactor fuel elements, and unconventional fuel elements are also covered. This volume consists of 12 chapters and begins with an overview of nuclear reactors and fuel elements, as well as fuel element design and development based on the reactor operator's approach, materials scientist's approach, and interdisciplinary approach. The reader is then introduced to different types of nuclear fuels and their irradiation behavior, considerations for using cladding and duct materials in fuel element design and development, and fuel element design and modeling. The chapters that follow focus on the testing of fuel element performance, experimental techniques and equipment for testing fuel element designs, and the performance of fuels for water reactors. Fuel elements for gas-cooled reactors, fast reactors, and research and test reactors are also described. The book concludes with an assessment of unconventional fuel elements. This book will be useful to fuel element technologists as well as materials scientists and engineers.
Author: J. H. Field Publisher: ISBN: Category : Fast reactors Languages : en Pages : 84
Book Description
An experimental program to evaluate the performance of FCR and EFCR fuel during transient operation is outlined, and the initial series of tests are described in some detail. Test results from five experiments in the TREAT reactor, using 1-in. OD SS-clad UO2 fuel specimens, are compared with regard to fuel temperatures, mechanical integrity, and post-irradiation appearance. Incipient fuel pin failure limits for transients are identified with maximum fuel temperatures in the range of 7000 deg F. Multiple transient damage to the cladding is likely for transients above the melting point of the fuel.
Author: NEA Nuclear Science Committee. Task Force on Scientific Issues Related to Fuel Behaviour Publisher: OECD ISBN: Category : Technology & Engineering Languages : en Pages : 96
Author: International Atomic Energy Agency Publisher: International Atomic Energy Agency ISBN: 9789201627094 Category : Technology & Engineering Languages : en Pages :
Book Description
This publication represents a summary and proceedings of two technical meetings on the subjects of 1) hot cell post-irradiation examination (PIE) techniques and poolside inspection of water reactor fuel assemblies, and 2) fuel rod instrumentation and in-pile measurement techniques. It provides an overview of the status of in-pile and post-irradiation techniques of water reactor fuel examination with an emphasis on experimental methods applied to high burnup fuel. Further important items addressed at these meetings were the information on progress achieved in non-destructive and destructive PIE techniques used for investigation of water reactor fuels, advanced PIE techniques such as 3-D tomography, as well as advanced in-pile measurement techniques for a better understanding of mechanisms of fuel behavior under irradiation.
Author: Publisher: ISBN: Category : Languages : en Pages :
Book Description
The thermal and pressure stresses in metallic plate and rod type nuclear fuel elements during a nuclear accident severe enough tc produce a metal-water reaction have been reviewed. The initial reactor pericds required to cause these severe accidents are in the range of 4 to 12 milliseconds. The result of the stress review presented in this report indicates that the metallic plate type fuel will swell from internal fission gas pressure, the rate of gas generation, and the magnitude and exposure to elevated central temperatures. Pulsed irradiation type experiments on sample fuel elements are desirable to determine the extent and consequences of this swelling including determination of the mechanism of fuel element failure and if dispersion occurs, measurement of the size of the particles. The stress review for the clad oxide rod type fuel, indicates that the oxide core will expand more than the Zircaloy cladding during severe reactor accidents producing cladding strains in the range of one to two percent. Experimentally, 7 to 10 percent strain is required to rupture irradiated Zircaloy tubing indicating that the cladding should not fail during the reactor transients considered. However, localized areas of the cladding may experience damage during fabrication, loss of ductility or strain concentrations sufficient to crack the cladding. There is no evidence to indicate that the cladding will be pulverized into small enough pieces to accelerate a metalwater reaction nor to modify the analytical models used to evaluate the core conditions during the accidents considered. Pulsed irradiation type experiments on fuel specimens simulating the conditions reviewed would be desirable to evaluate the transient material and mechanical properties of the fuel elements. (auth).
Author: Brian R. T. Frost
Author: Brian R. T. Frost
Author: J. H. Field
Author: Ralph P. Stein
Author: August W. Cronenberg
Author: NEA Nuclear Science Committee. Task Force on Scientific Issues Related to Fuel Behaviour
Author: International Atomic Energy Agency
Author: Howard S. Dreyer
Author: Richard L. Puthoff
Author: