Index Catalog // ScholarsArchive@OSU

Incorporating statistical uncertainty into nuclear material diversion pathway analysis using probabilistic risk assessment

Creator:: Pharn, Benjamin T.
Abstract:: A Level 1 probabilistic risk assessment (PRA) is applied to a plutonium button-making process, but in regards to the risk of diversion (or theft) of special nuclear material (SNM) rather than the risk of mechanical failure. The main objective of the project was to identify the relative vulnerability of areas...
Resource Type:: Masters Thesis
Full Text:: analysis by developing a quantitative measure for safeguards. The unavailability of data concerning the

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:: to measure the addition of fertile targets ................. 41 4.2: Detector properties required

Rapid establishment of emergency action areas as a consequence of large scale radioactive material releases from fixed nuclear facilities

Creator:: Fundak, Robert
Abstract:: Rapid emergency response decisions from a radiation release are necessary in order to prevent the general public from being exposed to a potential radiation hazard. A one meter exposure rate measurement is all that is necessary to establish relocation and food control areas when the ground is contaminated at the...
Resource Type:: Masters Thesis
Full Text:: deficiencies in commercial reactor safety were not hardware problems, they were management problems. TMI

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:: being prepared for safety analysis matrix testing in support of the NuScale Power Inc. (NSP) design

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

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:: Feedwater Control System 26 2.4.6 Safety Injection Control System 26 3. RETRAN Model of A PWR 251 3.1

SBLOCA analysis for nuclear plant shutdown operations

Creator:: Wang, Yi
Abstract:: A series of small break loss of coolant accident (SBLOCA) analyses in nuclear plant shutdown operations was simulated using the code RELAP5A,MOD3 version 8.0 to predict the SBLOCA phenomena in the Zion-l nuclear power plant The first objective is to study the impact of SBLOCA (1" and 2" breaks) on...
Resource Type:: Masters Thesis
Full Text:: Typical Westinghouse RHRS 4 2. Literature Review 8 2.1 Summary of Principal Safety Concerns 8 2.2

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:: for Privacy performance. It enhances the safety factor in the case of a loss of flow accident

Scaling analysis of the Direct Reactor Auxiliary Cooling System for Gas-cooled Fast Reactors during a Depressurized Loss of Forced Convection Event

Creator:: Blake, Grant C.
Abstract:: The Direct Reactor Auxiliary Cooling System (DRACS) is a passive safety system capable of removing decay heat directly from the reactor core. Its modularity makes it scalable for use in reactors with various power levels. Work has previously been completed to support inclusion of the DRACS in liquid metal reactors...
Resource Type:: Masters Thesis
Full Text:: Auxiliary Cooling System (DRACS) is a passive safety system capable of removing decay heat directly from

Probabilistic risk assessment of a vented fuel system utilized in a generation IV gas-cooled fast reactor nuclear power plant

Creator:: Lenhof, Renae
Abstract:: A limited scope Probabilistic Risk Assessment (PRA) of a vented nuclear fuel system for a Generation IV Gas-Cooled Fast Reactor Plant was performed. The goal of the study was to better understand the safety and licensing implications of vented fuel technology. A Level 1 PRA was performed to determine the...
Resource Type:: Masters Thesis
Full Text:: study was to better understand the safety and licensing implications of vented fuel technology. A

ScholarsArchive@OSU

Incorporating statistical uncertainty into nuclear material diversion pathway analysis using probabilistic risk assessment

Simulated antineutrino signatures of nuclear reactors for nonproliferation applications

Rapid establishment of emergency action areas as a consequence of large scale radioactive material releases from fixed nuclear facilities

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

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

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

SBLOCA analysis for nuclear plant shutdown operations

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

Scaling analysis of the Direct Reactor Auxiliary Cooling System for Gas-cooled Fast Reactors during a Depressurized Loss of Forced Convection Event

Probabilistic risk assessment of a vented fuel system utilized in a generation IV gas-cooled fast reactor nuclear power plant

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