Irradiation Behavior of Uranium Carbide Fuels PDF Download
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Author: D. I. Sinizer
Publisher:
ISBN:
Category : Nuclear fuels
Languages : en
Pages : 52
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Book Description
Author: D. I. Sinizer
Publisher:
ISBN:
Category : Nuclear fuels
Languages : en
Pages : 52
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Book Description
Author: J. A. L. Robertson
Publisher: New York : Gordon and Breach
ISBN:
Category : Technology & Engineering
Languages : en
Pages : 328
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Book Description
Author: Angel Madrid
Publisher:
ISBN:
Category : Liquid metal fast breeder reactors
Languages : en
Pages : 580
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Book Description
Author: J. V. Shennan
Publisher:
ISBN:
Category : Nuclear fuels
Languages : en
Pages : 36
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Book Description
Author:
Publisher: Springer Science & Business Media
ISBN: 3662060140
Category : Science
Languages : en
Pages : 371
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Book Description
The present volume A4 of the "Uranium" series of the Gmelin Handbook deals with two very important technological aspects of the nuclear fuel cycle: - the behavior of fuel elements during burnup in a nuclear reactor, and - the reprocessing of spent fuel to recover the non-fissioned uranium and newly created materials. The usefullifetime of a fuel element in a nuclear reactor depends strongly on the change of its chemical and physical properties during irradiation. Properties like thermal conductivity, swelling, creep, and oxygen-to-metal ratio are strongly affected by the intense neutron field and the energetic fission products. Furthermore, the high temperature gradient in a fuel element also produces alterations of the initial fuel. such as densification or U: Pu segregation. All of these effects are thoroughly discussed for the different kinds of fuels to be used in modern nuclear reactors today or in the future. The vast amount of very often Contradietory results in sometimes difficultly obtainable Iiterature has been summarized to create a compendium in this field with the two sections, on oxide and on carbide and nitride fuels, respectively. The chapters on reprocessing of spent fuels deal only with fuel elements of the uranium 235 thorium fuel cycle and with those containing fuel highly enriched in U. The treatment of U0 2 and (U,Pu)0 has already been given in the transuranic element series.
Author:
Publisher: Springer
ISBN: 9783662060162
Category : Science
Languages : en
Pages : 359
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Book Description
The present volume A4 of the "Uranium" series of the Gmelin Handbook deals with two very important technological aspects of the nuclear fuel cycle: - the behavior of fuel elements during burnup in a nuclear reactor, and - the reprocessing of spent fuel to recover the non-fissioned uranium and newly created materials. The usefullifetime of a fuel element in a nuclear reactor depends strongly on the change of its chemical and physical properties during irradiation. Properties like thermal conductivity, swelling, creep, and oxygen-to-metal ratio are strongly affected by the intense neutron field and the energetic fission products. Furthermore, the high temperature gradient in a fuel element also produces alterations of the initial fuel. such as densification or U: Pu segregation. All of these effects are thoroughly discussed for the different kinds of fuels to be used in modern nuclear reactors today or in the future. The vast amount of very often Contradietory results in sometimes difficultly obtainable Iiterature has been summarized to create a compendium in this field with the two sections, on oxide and on carbide and nitride fuels, respectively. The chapters on reprocessing of spent fuels deal only with fuel elements of the uranium 235 thorium fuel cycle and with those containing fuel highly enriched in U. The treatment of U0 2 and (U,Pu)0 has already been given in the transuranic element series.
Author: R. D. Leggett
Publisher:
ISBN:
Category : Uranium
Languages : en
Pages : 66
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Book Description
Author:
Publisher:
ISBN:
Category : Uranium as fuel
Languages : en
Pages : 56
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Book Description
Author: Christopher Matthews
Publisher:
ISBN:
Category : Bubbles
Languages : en
Pages : 183
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Book Description
The need for cheap reliable energy, while simultaneously avoiding uranium supply constraints makes uranium carbide (UC) fueled Gas Fast Reactors offer an attractive nuclear reactor design. In order to qualify the fuel, an enhanced understanding of the behavior of uranium carbide during operation is paramount. Due to a reduced re-solution rate, uranium carbide suffers from a buildup of very large fission gas bubbles. While these bubbles serve to reduce total fission gas release through the trapping of diffusing gas atoms, they lead to high swelling and ultimately dominate the microstructure of the fuel. The bubble size distribution is determined by the competing absorption rate and the rate of knock-out, or re-solution. As a result of the enhanced thermal dissipative properties of uranium carbide fuel, the atom-by-atom knockout process was shown to be an accurate representation of re-solution in uranium carbide. Furthermore, the Binary Collision Approximation was shown to appropriately model the re-solution event, bypassing computationally expensive Molecular Dynamics simulations. The code 3DOT was developed as an off-shoot of the code 3DTrim, both of which utilize the TRIM algorithm to calculate the kinematics of ions traveling through a material. Benefiting from modern methods and enhanced computational power, the model created in 3DOT results in a more fundamental understanding of the re-solution process in uranium carbide. A re-solution parameter that was an order of magnitude lower than previously determined was cal- culated in 3DOT. A decrease in the re-solution parameter as a function of radius occurred for low bubble radii, with a nearly constant re-solution parameter for bubble radii above 50 nm. Through comparative studies on the re-solution parameter for various values of implantation energy and atomic density in the bubble, we found that while the re-solution parameter did change slightly, the overall shape did not. A new application, BUCK, was built using the MOOSE framework to simulate the fission gas bubble concentration distribution. In order to build a bare-bones foundation, the simplistic yet historically prevalent physics that can be used to model fission gas bubble nucleation, growth, and knock-out were implemented as stepping stones until more advanced models for each physical process can be created. As the first step towards models that are based on first-principles, the new re-solution parameter was included and tested within BUCK. BUCK was tested using different parameters and behaved normally. However, from studies using representative simulation parameters, it is clear that the currently implemented theory does not adequately identify the growth mechanism that leads to larger bubbles. While this currently limits the applicability of BUCK in a full fuel pin calculation, it provides the baseline structure in which new physics can be implemented, and represents an important step towards understanding the complex behavior of fission gas bubbles.
Author: H. Holleck
Publisher:
ISBN: 9783662107188
Category : Chemistry
Languages : en
Pages : 281
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Book Description
The present volume Uranium C12 covers the binary and polynary carbides of uranium, including ternary carbides with nonmetals like the carbide oxides and carbide nitrides U(C, O) and U(C, N). The binary carbide UC and especially the mixed carbide (UO.80PUO.20)C are of special importance due to their potential as the fuel for advanced . Fast Breeder Reactors" because of properties such as the short doubling time, the high fissionable material density, and the good thermal conductivity. On the other hand, the dicarbide UC is of interest for . High 2 Temperature Reactors", especially in the form of the mixed carbide (U . Th .)C . For O80 o20 2 the first time, India used mixed uranium-plutonium carbide (U . PU .)C as the fuel for its O3 O7 own newly developed 15 MW Fast Breeder Reactor at Kalpakkam, south of Madras. el Because of the technological importance of the uranium carbides a lot of data were published only in reports. In most cases, it was the aim of these less-scientifically based studies to promote the carbide fuel development on an economical basis. The lack of analyti cal data on the purity of the samples, missing characterization of the present phases, etc., hQINever, does not allow the discussion of the results of such references in this handbook. Therefore, only reliable publications were cited. For the technical fabrication of uranium carbides and their irradiation behavior, see Volumes A3 and A4 of this Handbook.
