Index Catalog // ScholarsArchive@OSU

A comparative study of nodal course-mesh methods for pressurized water reactors

Creator:: Bukar, Kyari Abba
Abstract:: Several computer codes based on one and two-group diffusion theory models were developed for SHUFFLE. The programs were developed to calculate power distributions in a two-dimensional quarter core geometry of a pressurized power reactor. The various coarse-mesh numerical computations for the power calculations yield the following: the Borresen's scheme applied...
Resource Type:: Masters Thesis
Full Text:: power distribution of a reactor core model is necessary in any reactor analysis. In this chapter

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:: the nuclear power industry for safety analysis. They are used to model the effect that a transient

An analysis of the infinite periodic lattice approximation used in unit cell calculations

Creator:: Thomas, Mark Stephen
Abstract:: Unit cell or pin cell calculations form the basis for most nuclear core modeling. Because of this, it is of primary importance to perform these unit cell calculations accurately. The subject of this thesis is the analysis of one of the major approximations made in unit cell modeling, the infinite...
Resource Type:: Masters Thesis
Full Text:: . INTRODUCTION Nuclear reactor systems usually consist of an orderly array of fuel, moderator and control rods

Feed-and-bleed transient analysis of OSU APEX facility using the modern Code Scaling, Applicability, and Uncertainty method

Creator:: Hallee, Brian Todd
Abstract:: The nuclear industry has long relied upon bounding parametric analyses in predicting the safety margins of reactor designs undergoing design-basis accidents. These methods have been known to return highly-conservative results, limiting the operating conditions of the reactor. The Best-Estimate Plus Uncertainty (BEPU) method using a modernized version of the Code-Scaling,...
Resource Type:: Masters Thesis
Full Text:: extrapolating results to full-scale nuclear power plants. • Develop or utilize a pre-existing sampling method

Stability analysis of dynamic heat transfer

Creator:: Luo, Shyr Tung
Abstract:: The stability analysis for a heated tube system is an important safety feature of a nuclear power plant. Although the system theory is well established for linear systems described by ordinary differential equations, there is still a shortage of theory dealing with a distributed system, which is described by partial...
Resource Type:: Dissertation
Full Text:: from both a nuclear reactor and a steam generator. 4.1 MATHEMATICAL MODEL FOR REACTOR HEAT TRANSFER

Steady State Modeling of the Minimum Critical Core of the Transient Reactor Test Facility

Creator:: Alberti, Anthony L.
Abstract:: With the advent of next generation reactor systems and new fuel designs, the U.S. Department of Energy (DOE) has identified the need for the resumption of transient testing of nuclear fuels. The DOE has decided that the Transient Reactor Test Facility (TREAT) at Idaho National Laboratory (INL) is best suited...
Resource Type:: Masters Thesis
Full Text:: a root mean square, axially integrated radial power distribution difference of 0.22% compared to

Solution of the 2D quasi-diffusion low-order equations using a coupled nodal/finite volume discretization in Cartesian and hexagonal geometry

Creator:: Nes, Razvan
Abstract:: An important improvement in the area of reactor core neutronic modeling is the development and use of the methods based on "quasi-diffusion" (QD) low-order equations. This family of methods takes into account the transport exactly using "functionals" computed by solving transport equations, and is amenable to solution with a variety...
Resource Type:: Dissertation
Full Text:: plants or nuclear reactor at nuclear power plants. Fossil fuel plants combust coal, oil, or natural

An MCNP benchmark model of the large experimental ZPPR-19B sodium fast reactor

Creator:: Onufer, Mark
Abstract:: The International Handbook of Evaluated Reactor Physics Benchmark Experiments contains detailed descriptions of many different reactor facilities. A large portion of these experiments have not been fully modeled due to the unavailability of computational power at the time of the experiment’s execution. With the advent of renewed interest in Sodium...
Resource Type:: Masters Thesis
Full Text:: of Science in Nuclear Engineering presented on December 11, 2013. Title: An MCNP Benchmark Model

A RETRAN model for a hypothetical steam line break occurring at low power operation of the Trojan Nuclear Plant

Creator:: Jensen, Peter John
Abstract:: The goal of this study is the modeling of a postulated break of the steam line at the containment penetration of the Trojan Nuclear Plant which is owned and operated by Portland General Electric. To perform this modeling, the RETRAN computer code package was utilized. RETRAN is designed to provide...
Resource Type:: Masters Thesis
Full Text:: RETRAN Model for a Hypothetical Steam Line Break Occurring at Low Power Operation of the Trojan Nuclear

Numerical techniques for coupled neutronic/thermal hydraulic nuclear reactor calculations

Creator:: Betts, Curt M.
Abstract:: The solution of coupled neutronic/thermal hydraulic nuclear reactor calculations requires the treatment of the nonlinear feedback induced by the thermal hydraulic dependence of the neutron cross sections. As a result of these nonlinearities, current solution techniques often diverge during the iteration process. These instabilities arise due to the low level...
Resource Type:: Masters Thesis
Full Text:: , these cross sections, through the fission process, determine the power distribution within the reactor

ScholarsArchive@OSU

A comparative study of nodal course-mesh methods for pressurized water reactors

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

An analysis of the infinite periodic lattice approximation used in unit cell calculations

Feed-and-bleed transient analysis of OSU APEX facility using the modern Code Scaling, Applicability, and Uncertainty method

Stability analysis of dynamic heat transfer

Steady State Modeling of the Minimum Critical Core of the Transient Reactor Test Facility

Solution of the 2D quasi-diffusion low-order equations using a coupled nodal/finite volume discretization in Cartesian and hexagonal geometry

An MCNP benchmark model of the large experimental ZPPR-19B sodium fast reactor

A RETRAN model for a hypothetical steam line break occurring at low power operation of the Trojan Nuclear Plant

Numerical techniques for coupled neutronic/thermal hydraulic nuclear reactor calculations

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