Author: P.R. Rowland
Publisher:
ISBN:
Category :
Languages : en
Pages : 27
Book Description
Mechanisms of Fission Product Migration in Nuclear Graphite
Author: P.R. Rowland
Publisher:
ISBN:
Category :
Languages : en
Pages : 27
Book Description
Publisher:
ISBN:
Category :
Languages : en
Pages : 27
Book Description
Energy Research Abstracts
Author:
Publisher:
ISBN:
Category : Power resources
Languages : en
Pages : 600
Book Description
Publisher:
ISBN:
Category : Power resources
Languages : en
Pages : 600
Book Description
Diffusion of Fission Products in Nuclear Graphite
Author: Kevin Graydon
Publisher:
ISBN:
Category :
Languages : en
Pages : 55
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.
Publisher:
ISBN:
Category :
Languages : en
Pages : 55
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.
RATES AND MECHANISMS OF THE LOSS OF FISSION PRODUCTS FROM URANIUM-GRAPHITE FUEL MATERIALS.
Author:
Publisher:
ISBN:
Category :
Languages : en
Pages :
Book Description
Evidence concerning high-temperature loss of fissionproduct atoms from U- graphite fuel material is presented. Results indicate that no single model can adequately describe the behavior of all fission-product elements. Variables which affect loss rates are temperature, chemical species or element, sample dimensions, U concentration, fabrication, and graphite porosity. Loss-rate processes are discussed. (J.R.D.).
Publisher:
ISBN:
Category :
Languages : en
Pages :
Book Description
Evidence concerning high-temperature loss of fissionproduct atoms from U- graphite fuel material is presented. Results indicate that no single model can adequately describe the behavior of all fission-product elements. Variables which affect loss rates are temperature, chemical species or element, sample dimensions, U concentration, fabrication, and graphite porosity. Loss-rate processes are discussed. (J.R.D.).
Summary Report and New Proposal on the Mechanisms of Fission Gas Diffusion in Graphite
Author: W. S. Diethorn
Publisher:
ISBN:
Category : Diffusion
Languages : en
Pages : 66
Book Description
Publisher:
ISBN:
Category : Diffusion
Languages : en
Pages : 66
Book Description
Euro Abstracts
Author:
Publisher:
ISBN:
Category : Nuclear energy
Languages : en
Pages : 768
Book Description
Publisher:
ISBN:
Category : Nuclear energy
Languages : en
Pages : 768
Book Description
Fundamental Nuclear Energy Research
Author: U.S. Atomic Energy Commission. Division of Plans and Reports
Publisher:
ISBN:
Category : Nuclear energy
Languages : en
Pages : 434
Book Description
Publisher:
ISBN:
Category : Nuclear energy
Languages : en
Pages : 434
Book Description
Investigation of the Feasibility of Utilizing Gamma Emission Computed Tomography in Evaluating Fission Product Migration in Irradiated TRISO Fuel Experiments
Author:
Publisher:
ISBN:
Category :
Languages : en
Pages : 14
Book Description
In the High Temperature Gas-Cooled Reactor (HTGR) the TRISO particle fuel serves as the primary fission product containment. However the large number of TRISO particles present in proposed HTGRs dictates that there will be a small fraction (~10-4 to 10-5) of as manufactured and in-pile particle failures that will lead to some fission product release. The matrix material surrounding the TRISO particles in fuel compacts and the structural graphite holding the TRISO particles in place can also serve as sinks for containing any released fission products. However data on the migration of solid fission products through these materials is lacking. One of the primary goals of the AGR-3/4 experiment is to study fission product migration from failed TRISO particles in prototypic HTGR components such as structural graphite and compact matrix material. In this work, the potential for a Gamma Emission Computed Tomography (GECT) technique to non-destructively examine the fission product distribution in AGR-3/4 components and other irradiation experiments is explored. Specifically, the feasibility of using the Idaho National Laboratory (INL) Hot Fuels Examination Facility (HFEF) Precision Gamma Scanner (PGS) system for this GECT application is considered. To test the feasibility, the response of the PGS system to idealized fission product distributions has been simulated using Monte Carlo radiation transport simulations. Previous work that applied similar techniques during the AGR-1 experiment will also be discussed as well as planned uses for the GECT technique during the post irradiation examination of the AGR-2 experiment. The GECT technique has also been applied to other irradiated nuclear fuel systems that were currently available in the HFEF hot cell including oxide fuel pins, metallic fuel pins, and monolithic plate fuel.
Publisher:
ISBN:
Category :
Languages : en
Pages : 14
Book Description
In the High Temperature Gas-Cooled Reactor (HTGR) the TRISO particle fuel serves as the primary fission product containment. However the large number of TRISO particles present in proposed HTGRs dictates that there will be a small fraction (~10-4 to 10-5) of as manufactured and in-pile particle failures that will lead to some fission product release. The matrix material surrounding the TRISO particles in fuel compacts and the structural graphite holding the TRISO particles in place can also serve as sinks for containing any released fission products. However data on the migration of solid fission products through these materials is lacking. One of the primary goals of the AGR-3/4 experiment is to study fission product migration from failed TRISO particles in prototypic HTGR components such as structural graphite and compact matrix material. In this work, the potential for a Gamma Emission Computed Tomography (GECT) technique to non-destructively examine the fission product distribution in AGR-3/4 components and other irradiation experiments is explored. Specifically, the feasibility of using the Idaho National Laboratory (INL) Hot Fuels Examination Facility (HFEF) Precision Gamma Scanner (PGS) system for this GECT application is considered. To test the feasibility, the response of the PGS system to idealized fission product distributions has been simulated using Monte Carlo radiation transport simulations. Previous work that applied similar techniques during the AGR-1 experiment will also be discussed as well as planned uses for the GECT technique during the post irradiation examination of the AGR-2 experiment. The GECT technique has also been applied to other irradiated nuclear fuel systems that were currently available in the HFEF hot cell including oxide fuel pins, metallic fuel pins, and monolithic plate fuel.
Fission Product Holdup in Graphite. [HTGR].
Author:
Publisher:
ISBN:
Category :
Languages : en
Pages :
Book Description
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.
Publisher:
ISBN:
Category :
Languages : en
Pages :
Book Description
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.
Nuclear Science Abstracts
Author:
Publisher:
ISBN:
Category : Nuclear energy
Languages : en
Pages : 658
Book Description
Publisher:
ISBN:
Category : Nuclear energy
Languages : en
Pages : 658
Book Description