Index Catalog // ScholarsArchive@OSU

Absorption of methyl iodide by aqueous hydrazine solutions within spray chambers

Creator:: Postma, A. K.
Abstract:: Absorption of methyl iodide by aqueous hydrazine solutions within spray chambers was studied theoretically and experimentally. The objective of this work was to evaluate the absorption rate of airborne methyl iodide by sprays of hydrazine solutions under conditions expected to exist in post-accident nuclear reactor containment atmospheres. Experimental effort included...
Resource Type:: Dissertation
Full Text:: consequences in nuclear power plants is very low, safety facilities have been incorporated into reactor

Routine and post-accident analysis of nuclear power plant reactor coolant

Creator:: Jolma, Eva I.
Abstract:: The 1979 incident at Three Mile Island drew attention to weaknesses in the capability to obtain and analyze reactor coolant samples in an environment where the reactor core incurs significant damage. As a result, the U.S. Nuclear Regulatory Commission promulgated regulations requiring that reactor coolant analytical capabilities be upgraded so...
Resource Type:: Masters Thesis
Full Text:: research effort. Routine and Post-Accident Analysis of Nuclear Power Plant Reactor Coolant by Eva I

Chemical analysis for boron in nuclear reactor coolant water

Creator:: McDaniel, Dawn Marie
Resource Type:: Masters Thesis
Full Text:: : Chemical Analysis for Boron in Nuclear Reactor Coolant Water Signature redacted for privacy. John C

An investigative approach to study the RETRAN modeling concepts in the analysis of a safety valve discharge transient

Creator:: Youssefnia, Mohammad Hossein
Abstract:: This study focuses on investigation of the RETRAN modeling concepts in the analysis of a safety valve discharge transient. Three safety valve tests conducted at the Combustion Engineering test facility were chosen for evaluation with RETRAN. It was essential to apply the critical flow boundary condition at the downstream nodes...
Resource Type:: Dissertation
Full Text:: measure incorporated in the design of reactor safety systems, further pressure relief is also provided

Modeling transient thermalhydraulic behavior of a thermionic fuel element for nuclear space reactors

Creator:: Al-Kheliewi, Abdullah S.
Abstract:: A transient code (TFETC) for calculating the temperature distribution throughout the radial and axial positions of a thermionic fuel element (TFE) has been successfully developed. It accommodates the variations of temperatures, thermal power, electrical power, voltage, and current density throughout the TFE as a function of time as well as...
Resource Type:: Dissertation
Full Text:: Figure Page 1.1 SNAP-10 Nuclear Space Reactor 2 1.2 SP-100 Nuclear Space Reactor Configuration

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:: accident scenarios identified in the Oregon State TRIGA® Reactor (OSTR) Safety Analysis Report (SAR

Simulated antineutrino signatures of nuclear reactors for nonproliferation applications

Creator:: Misner, Alex C.
Abstract:: Antineutrino detectors could provide a valuable addition to current safeguards regimes. Antineutrinos are an attractive emission to monitor due to their low interaction cross-section that prevents them from being shielded and the dependence of their spectrum on the power level and isotopic content of a reactor core. While there are...
Resource Type:: Dissertation
Full Text:: reactor, such observations cannot predict the effect of the diversion of nuclear material on the

RELAP5-3D modeling of ADS blowdown of MASLWR facility

Creator:: Bowser, Christopher Jordan
Abstract:: Oregon State University has hosted an International Atomic Energy Agency (IAEA) International Collaborative Standard Problem (ICSP) through testing conducted on the Multi-Application Small Light Water (MASLWR) facility. The MASLWR facility features a full-time natural circulation loop in the primary vessel and a unique pressure suppression device for accident scenarios. Automatic...
Resource Type:: Masters Thesis
Full Text:: Reactor (MASLWR) facility at Oregon State University (OSU) which was funded by the Nuclear Energy

High pressure condensation heat transfer in the evacuated containment of a small modular reactor

Creator:: Casey, Jason R.
Abstract:: At Oregon State University the Multi-Application Small Light Water Reactor (MASLWR) integral effects testing facility is being prepared for safety analysis matrix testing in support of the NuScale Power Inc. (NSP) design certification progress. The facility will be used to simulate design basis accident performance of the reactor's safety systems....
Resource Type:: Masters Thesis
Full Text:: Nuclear Regulatory Commission SMR Small Modular Reactor NSSS Nuclear Steam Supply System LOCA Loss of

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

ScholarsArchive@OSU

Absorption of methyl iodide by aqueous hydrazine solutions within spray chambers

Routine and post-accident analysis of nuclear power plant reactor coolant

Chemical analysis for boron in nuclear reactor coolant water

An investigative approach to study the RETRAN modeling concepts in the analysis of a safety valve discharge transient

Modeling transient thermalhydraulic behavior of a thermionic fuel element for nuclear space reactors

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

Simulated antineutrino signatures of nuclear reactors for nonproliferation applications

RELAP5-3D modeling of ADS blowdown of MASLWR facility

High pressure condensation heat transfer in the evacuated containment of a small modular reactor

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

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