Index Catalog // ScholarsArchive@OSU

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

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:: coupled phenomena which occur within a nuclear reactor core, a great deal of mathematical modeling is

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

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:: that of an MCNP model for the Advanced Test Reactor (ATR) and INL have been completed. Attila

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:: nuclear reactor core analysis are (1) calculating flux distribution and (2) determining criticality

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

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

Nodal methods for calculating nuclear reactor transients, control rod patterns, and fuel pin powers

Creator:: Cho, Byung-Oh
Abstract:: Nodal methods which are used to calculate reactor transients, control rod patterns, and fuel pin powers are investigated. The 3-D nodal code, STORM, has been modified to perform these calculations. Several numerical examples lead to the following conclusions: (1) By employing a thermal leakage-to-absorption ratio11 (TLAR) approximation for the spatial...
Resource Type:: Dissertation

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:: in mathematical complexity. 1.2 Method of Analysis and Limitations The Symbolic Nuclear Analysis

Scaling of the Chinese HTR-PM Reactor Design for Licensing and Testing at the Oregon State University High Temperature Test Facility

Creator:: Cox, Jordan
Abstract:: The United States (US) Department of Energy (DOE) has identified the pebble-bed reactor as a high priority for US research with the end goal of licensing a pebble-bed reactor (PBR) for operation in the United States. Before this can be accomplished significant research must be done regarding the safety of...
Resource Type:: Masters Thesis
Full Text:: produced in a non-nuclear test facility rather than a nuclear reactor. Finally, in order to conduct the

ScholarsArchive@OSU

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

Numerical techniques for coupled neutronic/thermal hydraulic nuclear reactor calculations

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

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

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

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

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

Nodal methods for calculating nuclear reactor transients, control rod patterns, and fuel pin powers

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

Scaling of the Chinese HTR-PM Reactor Design for Licensing and Testing at the Oregon State University High Temperature Test Facility

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