Index Catalog // ScholarsArchive@OSU

Determination of a calculation bias in the MCNP model of the OSTR

Creator:: Kitto, Allyson K.
Abstract:: Oregon State University is home to a TRIGA® Mark II reactor. In October of 2008, the reactor began operating on low enriched uranium fuel. A model of the facility exists in MCNP, a Monte Carlo code that can be used for criticality calculations. Until now, a bias in the calculation...
Resource Type:: Masters Thesis
Full Text:: understood and used for future model improvement or manipulation. 1.2. OSU TRIGA Background The reactor

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:: Mark Stephen Thomas TABLE OF CONTENTS Chapter Page 1 INTRODUCTION 1 2 MATHEMATICAL THEORY 5 3

Scaling analysis for the pebble bed of the very high temperature gas-cooled reactor thermal hydraulic test facility

Creator:: Nelson, Benjamin L.
Abstract:: The Very High Temperature Reactor (VHTR) has two possible core configurations, a hexagonal prismatic and a pebble bed. It is essential that an experimental facility be built for the validation of computer codes for the safe operation of the VHTR. The scaling of the prismatic core configuration has been analyzed...
Resource Type:: Masters Thesis
Full Text:: -cooled Reactor Thermal Hydraulic Test Facility Abstract approved

Applying uncertainty and sensitivity on thermal hydraulic subchannel analysis for the multi-application small light water reactor

Creator:: Brigantic, Adam
Abstract:: Small modular reactors (SMRs) are a recent advancement in commercial nuclear reactor design with growing interest worldwide. New SMR concepts, such as the Multi-Application Small Light Water Reactor (MASLWR), must undergo a licensing processes established by the U.S. Nuclear Regulatory Commission (NRC) prior to commercial operation. Given the lack of...
Resource Type:: Masters Thesis
Full Text:: mathematical system model. Epistemic uncertainty results from the inability to know an appropriate value for

Simulation of a membrane reactor for ammonia decomposition

Creator:: Kraisuwansarn, Nichakorn
Abstract:: The purpose of this research is to study the feasibility of combining reaction and separation in a membrane separation unit. The asymmetric ceramic membrane reactor applied to the gas-phase catalytic decomposition reaction of ammonia from an IGCC (Integrated Gasification Combined Cycle Processes) gas mixture was simulated in the temperature range...
Resource Type:: Masters Thesis

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

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:: bounding parametric analyses in predicting the safety margins of reactor designs undergoing design-basis

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:: mathematical models implemented in computer codes. These codes simulate the neutron behavior in the reactor

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:: experimental facility in order to determine any bias in the MCNP model. These results are useful to reactor

Dynamic simulation of a Westinghouse PWR four-loop plant during steam generator tube rupture event

Creator:: Sui, Yueh Chun
Abstract:: This study addresses the application of a full reactor system computer code to simulate the reactor primary and secondary loop thermal hydraulics, plant control and protection systems during the steam generator tube rupture (SGTR) events. The Trojan nuclear plant which is owned and operated by the Portland General Electric (PGE)...
Resource Type:: Masters Thesis
Full Text:: Abstract Approved :, S. Anghaie This study addresses the application of a full reactor system computer

ScholarsArchive@OSU

Determination of a calculation bias in the MCNP model of the OSTR

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

Scaling analysis for the pebble bed of the very high temperature gas-cooled reactor thermal hydraulic test facility

Applying uncertainty and sensitivity on thermal hydraulic subchannel analysis for the multi-application small light water reactor

Simulation of a membrane reactor for ammonia decomposition

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

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

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

Dynamic simulation of a Westinghouse PWR four-loop plant during steam generator tube rupture event

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