Diffusion of Fission Products in Nuclear Graphite PDF Download
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Author: Kevin Graydon
Publisher:
ISBN:
Category :
Languages : en
Pages : 55
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Book Description
The next generation of nuclear reactors (Generation IV) are specified to use graphite as the choice material for neutron moderation and structural components. For this reason, study of diffusion behavior of the fission product, ruthenium (Ru), in the candidate graphite grades, POCO AXF-5Q & ZXF-5Q, IG-110, NBG-18 and PCEA, is necessary. Diffusion data for Ru in these grades is absent due to the lack of prior studies and this work aims to fill this void and begin the study of this metallic fission product’s diffusion behavior. By utilizing physical vapor deposition (PVD), dynamic secondary ion mass spectroscopy (SIMS), and the thin film solution to the diffusion equation coupled with a short circuit diffusion correction term, the diffusivities and Arrhenius temperature dependence are determined in the range of 500-1000°C and reported for the first time. Diffusion of Ru in this range is slow, with both lower activation energies and diffusion coefficients than other metallic fission products. Simulations of Ru diffusion on a graphene plane give activation energies similar to those acquired in this study. This is consistent with Ru behaving as an intercalating species and using the region in between graphene-like basal planes of graphite to travel through the lattice.
Author: Kevin Graydon
Publisher:
ISBN:
Category :
Languages : en
Pages : 55
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Book Description
The next generation of nuclear reactors (Generation IV) are specified to use graphite as the choice material for neutron moderation and structural components. For this reason, study of diffusion behavior of the fission product, ruthenium (Ru), in the candidate graphite grades, POCO AXF-5Q & ZXF-5Q, IG-110, NBG-18 and PCEA, is necessary. Diffusion data for Ru in these grades is absent due to the lack of prior studies and this work aims to fill this void and begin the study of this metallic fission product’s diffusion behavior. By utilizing physical vapor deposition (PVD), dynamic secondary ion mass spectroscopy (SIMS), and the thin film solution to the diffusion equation coupled with a short circuit diffusion correction term, the diffusivities and Arrhenius temperature dependence are determined in the range of 500-1000°C and reported for the first time. Diffusion of Ru in this range is slow, with both lower activation energies and diffusion coefficients than other metallic fission products. Simulations of Ru diffusion on a graphene plane give activation energies similar to those acquired in this study. This is consistent with Ru behaving as an intercalating species and using the region in between graphene-like basal planes of graphite to travel through the lattice.
Author: J. Bromley
Publisher:
ISBN:
Category :
Languages : en
Pages : 46
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Author: Lukas M. Carter
Publisher:
ISBN:
Category :
Languages : en
Pages : 190
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Book Description
Release of radioactive fission products from nuclear fuel during normal reactor operation or in accident scenarios is a fundamental safety concern. Of paramount importance are the understanding and elucidation of mechanisms of chemical interaction, nuclear interaction, and transport phenomena involving fission products. Worldwide efforts to reduce fossil fuel dependence coupled with an increasing overall energy demand have generated renewed enthusiasm toward nuclear power technologies, and as such, these mechanisms continue to be the subjects of vigorous research. High-Temperature Gas-Cooled Reactors (HTGRs or VHTRs) remain one of the most promising candidates for the next generation of nuclear power reactors. An extant knowledge gap specific to HTGR technology derives from an incomplete understanding of fission product transport in major core materials under HTGR operational conditions. Our specific interest in the current work is diffusion in reactor graphite. Development of methods for analysis of diffusion of multiple fission products is key to providing accurate models for fission product release from HTGR core components and the reactor as a whole. In the present work, a specialized diffusion cell has been developed and constructed to facilitate real-time diffusion measurements via ICP-MS. The cell utilizes a helium gas-jet system which transports diffusing fission products to the mass spectrometer using carbon nanoparticles. The setup was designed to replicate conditions present in a functioning HTGR, and can be configured for real-time release or permeation measurements of single or multiple fission products from graphite or other core materials. In the present work, we have analyzed release rates of cesium in graphite grades IG-110, NBG-18, and a commercial grade of graphite, as well as release of iodine in IG-110. Additionally we have investigated infusion of graphite samples with Cs, I, Sr, Ag, and other surrogate fission products for use in release or profile measurements of diffusion coefficients.
Author: Seishi Yajima
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ISBN:
Category :
Languages : en
Pages : 0
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Author: P.R. Rowland
Publisher:
ISBN:
Category :
Languages : en
Pages : 27
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Author: R. E. Nightingale
Publisher: Academic Press
ISBN: 1483258483
Category : Technology & Engineering
Languages : en
Pages : 566
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Book Description
Nuclear Graphite focuses on the development and uses of nuclear graphite, including machining practices, manufacture, nuclear properties and structure, radiation, and electrical resistance. The selection first discusses the applications of graphite in the nuclear industry, machining practices, and manufacture. Discussions focus on early, current, and future applications of graphite, impregnation, graphitization, purification, general machining techniques, and equipment and methods in the nuclear industry. The book then examines the structure and nuclear and properties of graphite. The text evaluates radiation-induced structural and dimensional changes; radiation effects on electrical and thermal properties; and radiation effects on mechanical properties. Topics include radiation effects on crystal structure, electrical resistance, thermoelectric power, magnetoresistance, coefficient of friction, irradiation under stress, and elastic moduli of nuclear graphite. The book also ponders on stored energy, annealing radiation effects, and gas-graphite systems. The selection is a dependable source of data for readers interested in the applications of nuclear graphite.
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Languages : en
Pages :
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Author:
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Languages : en
Pages :
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The transport of fission products from points of origin in unclad graphite matrix-type fuels to the reactor circulating system involves, as one of the steps. diffusion through the graphite matrix to the fuel element surface. As pointed out by Rosenthal. the fraction of a given fission product chain actaally reaching the fuel element surface will be small if the time for transport through the graphite is long compared to the half-lives of the volatile members. An important problem, therefore, is the determination of the effective transport rates of the various mobile elements and their daughter products of interest through various graphites suitable for use as fuel element compacts. as functions of temperature over the range of greatest immediate interest to reactor designers. The upper end of the range need not exceed about 1000 deg C. The transport of helium and arbon through various graphites has been the subject of considerable study by Watson. Evans, and other, . and a prelimilnary investigation of the high temperature transport of some ordinarily non-volatile elements has been carried out by Saunders. This work is briefly reviewed in relation to the final problem and the areas in which further information is needed most by reactor designers is indicated. (auth).
Author:
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Category :
Languages : en
Pages :
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The goal of this project is to determine changes in adsorption and desorption of fission products to/from nuclear-grade graphite in response to a changing chemical environment. First, the project team will employ principle calculations and thermodynamic analysis to predict stability of fission products on graphite in the presence of structural defects commonly observed in very high- temperature reactor (VHTR) graphites. Desorption rates will be determined as a function of partial pressure of oxygen and iodine, relative humidity, and temperature. They will then carry out experimental characterization to determine the statistical distribution of structural features. This structural information will yield distributions of binding sites to be used as an input for a sorption model. Sorption isotherms calculated under this project will contribute to understanding of the physical bases of the source terms that are used in higher-level codes that model fission product transport and retention in graphite. The project will include the following tasks: Perform structural characterization of the VHTR graphite to determine crystallographic phases, defect structures and their distribution, volume fraction of coke, and amount of sp2 versus sp3 bonding. This information will be used as guidance for ab initio modeling and as input for sorptivity models; Perform ab initio calculations of binding energies to determine stability of fission products on the different sorption sites present in nuclear graphite microstructures. The project will use density functional theory (DFT) methods to calculate binding energies in vacuum and in oxidizing environments. The team will also calculate stability of iodine complexes with fission products on graphite sorption sites; Model graphite sorption isotherms to quantify concentration of fission products in graphite. The binding energies will be combined with a Langmuir isotherm statistical model to predict the sorbed concentration of fission products on each type of graphite site. The model will include multiple simultaneous adsorbing species, which will allow for competitive adsorption effects between different fission product species and O and OH (for modeling accident conditions).
![Fission Product Holdup in Graphite. [HTGR]. PDF](https://books.google.com/books/content/images/frontcover/cUS1AQAACAAJ?fife=w80-h100)
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Languages : en
Pages :
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Multicomponent time-dependent concentration diffusion and radioactive decay of isotopic species is an important aspect of fission product migration and release from fuel particles and fuel elements in a High Temperature Gas-Cooled Reactor (HTGR). After fission products escape from a fuel particle in an HTGR, it is still necessary for them to diffuse across the graphite web of a fuel block to a coolant hole before they can be entrained in the primary coolant. The time required for a given fission product species to diffuse across the graphite web has a direct influence on the time-dependent release associated with a significant increase in the power/flow ratio. The main purpose of the paper is to present the results of a study of the holdup time in graphite of Sr as a function of the diffusion constants. The study employs a newly-developed multicomponent time-dependent diffusion and decay code called DASH. Analysis methods for solving the type of problem discussed are well known, and some applications to fission product decay and diffusion in HTGRs have appeared in the literature. However, the methods employed are often subject to time step limitations, and the effects of decay are not adequately handled. The DASH code uses a one dimensional spatial discretization for the diffusion operator and an analytic matrix operator method to remove the time dependence. Comparisons of the solutions given by DASH with a number of analytic solutions have been made, and in all instances considered the agreement with analytical solutions is excellent and limited only by the inaccuracy inherent in the spatial discretization.
