Index Catalog // ScholarsArchive@OSU

Application of sandwich structure analysis in predicting critical flow velocity for a laminated flat plate

Creator:: Jensen, Philip
Abstract:: The Oregon State University (OSU), Hydro Mechanical Fuel test Facility (HMFTF) is designed to hydro-mechanically test prototypical plate type fuel. OSU's fuel test program is a part of the Global Threat Reduction Initiative (GTRI), formerly known as the Reduced Enrichment for Research and Test Reactor program. One of the GTRI's...
Resource Type:: Masters Thesis
Full Text:: enriched uranium (HEU) to a low enriched uranium (LEU) fuel in an effort to reduce nuclear proliferation

Vibration Analysis of Advanced Test Reactor Miniplate Hydraulic Test in the OSU HMFTF

Creator:: Latimer, Griffen D.
Abstract:: Current research on the topic of advanced reactor fuel types include the use of ultrahigh density Uranium-Molybdenum fuels, towards their use in high-performance research reactors. These reactors operate with high power densities, and the increased cooling requirements therefore place high relevance on the fluid-structure interaction with these fuel elements; therefore...
Resource Type:: Masters Thesis
Full Text:: conveying fluid. Numerous studies have investigated the dynamic response of shells under these conditions

A study of opposing mixed convection in the GRTS and in downward pipe flows using the FLUENT CFD code

Creator:: Jackson, R. Brian
Abstract:: Downward flow opposing mixed convection conditions have been studied using the FLUENT CFD code. This work was comprised of two primary objectives. The first was to study core channel flow reversal susceptibility in the Gas Reactor Test Section (GRTS) at low flow conditions. The second objective was the development of...
Resource Type:: Masters Thesis
Full Text:: Jackson for the degree of Master of Science in Nuclear Engineering presented on March 17, 2008

Pressure losses across a 180 degree turn in the AP1000 cooling annulus

Creator:: Nelson, Roy K.
Abstract:: A one-sixth scaled facility was constructed at Oregon State University to determine the pressure losses around Westinghouse's AP1000 Passive Containment Cooling System. An explanation of the AP1000 Passive Containment Cooling System is given. This study examined the pressure loss and loss coefficient around the 180 degree turn of the Passive...
Resource Type:: Masters Thesis
Full Text:: Master of Science in Nuclear Engineering presented on October 2, 2008. Title: Pressure Losses Across a

Comparative analysis of the Žukauskas method and data from the OSU MASLWR test facility steam generator

Creator:: Wuth, Bradyn J.
Abstract:: Some of the next generation of small modular reactors have been designed with helically coiled steam generators integrated into the reactor vessel to enhance natural circulation. According to the developers of TASS/SMR, a one-dimensional thermal hydraulic code written to model the Korean Atomic Energy Research Institute's SMART reactor, the Žukauskas...
Resource Type:: Masters Thesis
Full Text:: to a nuclear core's fuel assemblies. The lower core plate and shroud around the core keep the

An Approach to Modeling Fuel Plate Deformations in Fluid-Structure Interactions

Creator:: Howard, Trevor
Abstract:: As computational power increases, so does the desire to use fluid-structure interaction (FSI) software to design complex systems and components such as fuel plates. Presently, an effort is underway to support the design and qualification of a new nuclear fuel material which is intended for applications in select, plate-type, research...
Resource Type:: Masters Thesis
Full Text:: An Approach to Modeling Fuel Plate Deformations in Fluid

A thermal hydraulics analysis of a molybdenum production element for implementation in TRIGA reactors

Creator:: Byfield, Patrick
Abstract:: The metastable isotope of technetium-99 (Tc-99m) is an important diagnostic tool used in the field of nuclear medicine due to the isotope's 6.0 hour half-life, 140.5 keV γ-decay mechanism, and multiple oxidation states [1,2]. Approximately 70% of the world’s nuclear medicine procedures involve the use of Tc-99m [3]. The conventional...
Resource Type:: Masters Thesis
Full Text:: , wherein the fluid is heated by fuel rods near the bottom of the reactor pool and cooling is provided by

Thermal hydraulic analysis of the Oregon State TRIGA Reactor using RELAP5-3D

Creator:: Marcum, Wade R.
Abstract:: Oregon State University has recently conducted a complete core conversion analysis as part of the Reduced Enrichment for Research and Test Reactors Pprogram. The goals of the thermal hydraulic analyses were to calculate natural circulation flow rates, coolant temperatures and fuel temperatures as a function of core power for both...
Resource Type:: Masters Thesis
Full Text:: Wade R. Marcum for the degree of Master of Science in Nuclear Engineering presented on January 23

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:: likelihood of a nuclear accident using fuel design and material properties. To this end this company began

Nitrogen concentration sensitivity study of the lock exchange phenomenon in the high temperature test facility

Creator:: Utberg, John E., Jr
Abstract:: The Modular High Temperature Gas-Cooled Reactor (MHTGR) is a graphite moderated reactor that utilizes helium as its coolant. One consideration of importance is how the MHTGR will perform during a Depressurized Conduction Cooldown (DCC) accident, which generally can be divided into three phases: depressurization, air ingress, and natural circulation. After...
Resource Type:: Masters Thesis
Full Text:: codes. Although the HTTF is a scaled-down model of a nuclear reactor, there is no nuclear fuel in the

ScholarsArchive@OSU

Application of sandwich structure analysis in predicting critical flow velocity for a laminated flat plate

Vibration Analysis of Advanced Test Reactor Miniplate Hydraulic Test in the OSU HMFTF

A study of opposing mixed convection in the GRTS and in downward pipe flows using the FLUENT CFD code

Pressure losses across a 180 degree turn in the AP1000 cooling annulus

Comparative analysis of the Žukauskas method and data from the OSU MASLWR test facility steam generator

An Approach to Modeling Fuel Plate Deformations in Fluid-Structure Interactions

A thermal hydraulics analysis of a molybdenum production element for implementation in TRIGA reactors

Thermal hydraulic analysis of the Oregon State TRIGA Reactor using RELAP5-3D

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

Nitrogen concentration sensitivity study of the lock exchange phenomenon in the high temperature test facility

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