Index Catalog // ScholarsArchive@OSU

A methodology for safeguards evaluation and its application to mixed oxide fuel fabrication

Creator:: Azarfar, Mohammad Ali
Abstract:: This study focuses on the development of a methodology to evaluate safeguardability of various processes in the nuclear fuel cycle. National and international implementation of safeguards systems has been investigated. It was concluded that proliferation is purely a political question that, hopefully, can be prevented by international norms and diplomacy;...
Resource Type:: Dissertation
Full Text:: technical differences between the different processes. On the other hand, the prevention of diversion does

Quantifying System Distortions between the VTR and IVTR facilities through H2TS Scaling Methodology and RELAP5-3D Simulations

Creator:: Johnson, Juwan A.
Abstract:: Sodium Cooled Fast Reactors (SFR) are complex and expensive due to high operating temperatures and neutron flux which pose technical challenges for these reactors. To overcome these challenges with SFRs, research and development of small scale and large-scale facilities are needed to bridge the gap in understanding. Scaling analyses develop...
Resource Type:: Masters Thesis
Full Text:: operating temperatures and neutron flux which pose technical challenges for these reactors. To overcome

Detection of actinides via nuclear isomer de-excitation

Creator:: Francy, Christopher J.
Abstract:: The detection of illicit weapons-grade nuclear material is a significant technical challenge receiving a great deal of attention in the nuclear community. A proposed solution is to actively bombard the material under investigation with neutrons, populating nuclear isomers in any potential fissile material. These isomers decay quickly, releasing gamma rays...
Resource Type:: Masters Thesis
Full Text:: technical challenge receiving a great deal of attention in the nuclear community. A proposed solution is

Critical heat flux estimation for annular channel geometry

Creator:: Pagh, Richard T.
Abstract:: Critical Heat Flux (CHF) is an important safety parameter for the design of nuclear reactors. The most commonly used predictive tool for determination of CHF is a look-up table developed using tube data with an average hydraulic test diameter of 8 mm. There exist in the world today nuclear reactors...
Resource Type:: Dissertation
Full Text:: annular channels, thus questioning its ability to perform correct safety evaluations. Since no better

Out-of-pile blowdown critical heat flux experiment

Creator:: Moussaoui, Musa A.
Abstract:: In the most challenging nuclear power plant accidents, transient critical heat flux (CHF) is a primary phenomenon that drives peak cladding temperature and ultimately fuel failure. It is not yet determined whether the use of steady-state CHF methods can accurately predict transient CHF under the conditions of a blowdown due...
Resource Type:: Dissertation
Full Text:: acceptable results. Moreover, there was no significant correlation between prediction accuracy and

A feasibility analysis of high conversion ratio pressurized water reactor designs

Creator:: Kliewer, Charles P.
Abstract:: A significant amount of interest has been aroused recently concerning the advancement of the current pressurized water reactor core designs with a special emphasis towards improving the conversion characteristics of these reactors. Most reports have been divided into two camps; those that deal with the neutronic aspects and those that...
Resource Type:: Masters Thesis
Full Text:: with only configurational changes in the core itself and no major equipment changes. Hence the author

A study of the behavior of actinides continuously recycled in a hard spectrum reactor

Creator:: Schofield, Paul Edwin
Resource Type:: Masters Thesis
Full Text:: 1000 years. Isolation for such a time period is not a serious technical problem within a 3 stable

Verification & Analysis of RELAP5-3D Code Modeling Diffusive Air Ingress Methods and Their Effects on Natural Circulation Time

Creator:: Schlamp, Nicklaus K.
Abstract:: The Energy Policy Act of 2005 dictates that the Nuclear Regulatory Commission (NRC) in conjunction with the Department of Energy (DOE) are responsible for the licensing and regulation safety strategies of the Very High Temperature Gas-cooled Reactor (VHTGR). The VHTGR is a next generation nuclear design that is capable of...
Resource Type:: Masters Thesis
Full Text:: analysis compared to Computational Fluid Dynamic (CFD) programs. No studies have been done to determine if

Investigation of the IRWST flow patterns during a simulated station blackout experiment on the OSU APEX facility

Creator:: Strohecker, Mark F.
Abstract:: The OSU/APEX thermal hydraulic test facility models the passive safety systems of the Westinghouse AP600 advanced light water reactor design. Numerous experiments have been performed to test these systems, the one of focus here is the station blackout scenario. This experiment simulated the complete loss of AC power to all...
Resource Type:: Masters Thesis
Full Text:: stratification was observed in the IRWST with no significant thermal mixing. A thermal layer developed in the

Feasibility of GOTHIC for a Pressurized Conduction Cooldown Event in the High Temperature Test Facility

Creator:: Brawley, Luke P.
Abstract:: The purpose of this work was to explore the feasibility of using the GOTHIC thermal hydraulics software package to model a pressurized conduction cooldown event in the High Temperature Test Facility. To achieve that aim, a model of the facility was constructed in the GOTHIC software and a selection of...
Resource Type:: Masters Thesis
Full Text:: results of the GOTHIC model can be compared against experimental data, no general conclusions can be

ScholarsArchive@OSU

A methodology for safeguards evaluation and its application to mixed oxide fuel fabrication

Quantifying System Distortions between the VTR and IVTR facilities through H2TS Scaling Methodology and RELAP5-3D Simulations

Detection of actinides via nuclear isomer de-excitation

Critical heat flux estimation for annular channel geometry

Out-of-pile blowdown critical heat flux experiment

A feasibility analysis of high conversion ratio pressurized water reactor designs

A study of the behavior of actinides continuously recycled in a hard spectrum reactor

Verification & Analysis of RELAP5-3D Code Modeling Diffusive Air Ingress Methods and Their Effects on Natural Circulation Time

Investigation of the IRWST flow patterns during a simulated station blackout experiment on the OSU APEX facility

Feasibility of GOTHIC for a Pressurized Conduction Cooldown Event in the High Temperature Test Facility

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