Index Catalog // ScholarsArchive@OSU

Fuel-cladding mechanical interaction in simulated sphere pac fuel

Creator:: Vincent, Andrew M.
Abstract:: A study has been made to investigate the fuel-cladding mechanical interaction (FCMI) in sphere pac fuels. An FCMI simulation has been constructed using the JVG Apparatus. This model has been used to investigate the mechanical interaction between a sphere pac bed and a cladding wall in terms of the cladding...
Resource Type:: Masters Thesis
Full Text:: Nuclear Engineering presented on Dec. 9, 1981 Title: Fuel-Cladding Mechanical Interaction In Simulated

Characterization of Oxide Growth of Aluminum Alloy 6061 for Nuclear Research and Test Reactors

Creator:: Wei, Youran
Abstract:: Oxide growth is a common phenomenon on aluminum alloy cladding surfaces in nuclear research and test reactors. Without better understanding of oxide growth, excessive oxide build up on the cladding surface reduces heat transfer capability from fuel to coolant and causes safety concerns due to the potential cladding rupture or...
Resource Type:: Masters Thesis
Full Text:: cladding surfaces in nuclear research and test reactors. Without better understanding of oxide growth

A model for the mechanical behavior of irradiated mixed carbide sphere pac fuel pins

Creator:: George, Thomas L.
Abstract:: A model is developed that allows the prediction of the stress-strain distribution in a mixed carbide sphere pac fuel pin during irradiation. The model is coupled with a previously developed model for the thermal behavior of the pin. The one-dimensional (radial at various axial locations) solution is for quasi-steady pin...
Resource Type:: Dissertation
Full Text:: the fuel and cladding separately so that slip at the fuel-cladding interface can be accommodated

A model for waterside oxidation of zircaloy fuel cladding in pressurized water reactors

Creator:: Almarshad, Abdullah I. A.
Abstract:: A model has been developed to predict the long-term oxidation rate of Zircaloy-4 for ex-reactor (autoclave) and in-reactor (PWR) environments and operating conditions. A computer program has been written to solve the oxygen diffusion equation by employing a fully implicit finite difference method for a one dimensional cylindrical geometry. The...
Resource Type:: Dissertation
Full Text:: A model for waterside oxidation of zircaloy fuel cladding

Optimization of in-core nuclear fuel management in a pressurized water reactor

Creator:: Stout, Richard Bartholomew
Abstract:: Fuel loading patterns which have a minimum power peak are economically desirable to allow power reactors to operate at the highest possible power density and to minimize the possibility of fuel failure. A computer code called SHUFLE was developed for pressurized water reactors which shuffles the fuel in search of...
Resource Type:: Dissertation
Full Text:: : OPTIMIZATION OF IN-CORE NUCLEAR FUEL MANAGE- MENT IN A PRESSURIZED WATER REACTOR Redacted for privacy

Modeling the thermal effects of fabricated fuel-to-cladding gap in fast reactor fuel

Creator:: Baker, Ronald Blair
Abstract:: This report presents a refined method for the calculation of fuel centerline temperatures and the linear-heat-rating-to-incipient melting, ' Q'[subscript m], of mixed-oxide fuel typical of that to be used in the Fast Flux Test Facility and in Liquid Metal Fast Breeder Reactors. Of primary concern are the effects of the...
Resource Type:: Masters Thesis
Full Text:: -cladding gap in fast reactor fuel AN ABSTRACT OF THE THESIS OF Ronald Blair Baker for the degree of

An assessment of describing function methods applied to reactor fuel performance analysis

Creator:: Nielsen, Larry H.
Abstract:: Large digital computer reactor fuel performance codes perform repeated calculations of temperature profiles during transients. The thermal response is required to calculate the mechanical behavior. This mechanical response in turn feeds back to the thermal calculations through mechanisms such as fuel swelling or shrinking, causing changes in the fuel-clad gap...
Resource Type:: Masters Thesis
Full Text:: reactor fuel performance analysis AN ABSTRACT OF THE THESIS OF Larry H. Nielsen for the degree of

The applicability of sintering mechanisms in the initial stage restructuring and preferential pore migration of irradiated (U₀.₈₅Pu₀.₁₅)C sphere-pac fuel

Creator:: Guenther, Richard John
Resource Type:: Masters Thesis
Full Text:: (U0.85Pu0.15)C spheres contained in typical fuel cladding. Mixed carbides such as (U0.85 Pu 0.15 )C have the

Stochastic simulation of sphere beds and the effect of packing density on the thermal conductivity of sphere pac nuclear fuels

Creator:: Henke, Robert K.
Abstract:: The establishment of any thermal model of reactor fuel requires a thorough understanding of the behavior of the thermal conductivity. This is especially true in the case of sphere pac fuels where the conductivity is a function of many variables beyond those normally affecting conventional fuels. One of these variables,...
Resource Type:: Masters Thesis
Full Text:: Fraction Sphere Pac Nuclear Fuel 42 III.lb Comparison of Experimental and Predicted Values of Thermal

Thermal and mechanical behavior of sphere pac fuel

Creator:: Montgomery, Scott D.
Abstract:: The use of advanced concepts can make the nuclear fuel cycle more efficient by allowing pins to be operated at high powers and to higher burnups. Among the more promising concepts in advanced fuels is sphere pac fuel. In a first attempt to bring together models describing individual phenomena, the...
Resource Type:: Masters Thesis
Full Text:: approved: Redacted for Privacy K1. L. Peiddicord The use of advanced concepts can make the nuclear fuel

ScholarsArchive@OSU

Fuel-cladding mechanical interaction in simulated sphere pac fuel

Characterization of Oxide Growth of Aluminum Alloy 6061 for Nuclear Research and Test Reactors

A model for the mechanical behavior of irradiated mixed carbide sphere pac fuel pins

A model for waterside oxidation of zircaloy fuel cladding in pressurized water reactors

Optimization of in-core nuclear fuel management in a pressurized water reactor

Modeling the thermal effects of fabricated fuel-to-cladding gap in fast reactor fuel

An assessment of describing function methods applied to reactor fuel performance analysis

The applicability of sintering mechanisms in the initial stage restructuring and preferential pore migration of irradiated (U₀.₈₅Pu₀.₁₅)C sphere-pac fuel

Stochastic simulation of sphere beds and the effect of packing density on the thermal conductivity of sphere pac nuclear fuels

Thermal and mechanical behavior of sphere pac fuel

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