Index Catalog // ScholarsArchive@OSU

Cost-Benefit Analysis : Proposed Safety Upgrades to Currently Operating Nuclear Power Plants

Creator:: Jordahl, Jonathan T.
Abstract:: In the wake of nuclear accidents such as Three Mile Island Unit 2 and Fukushima, the nuclear power industry’s safety record is scrutinized. Today the main concerns lies with hydrogen production in a nuclear reactor core when the zirconium fuel cladding reacts with the water coolant during an accident, creating...
Resource Type:: Masters Thesis
Full Text:: Engineering presented on September 13, 2016. Title: Cost-Benefit Analysis: Proposed Safety Upgrades to

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

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

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:: Utilized in a Generation IV Gas-Cooled Fast Reactor Nuclear Power Plant AN ABSTRACT OF THE

Flow Dynamics and Condensation of Film Flows in Small Modular Reactors

Creator:: Lee, Dongyoung
Abstract:: There is renewed interest in the reliability and safety of nuclear power plants following the Fukushima Daiichi nuclear accident followed by 8.9 magnitude earthquake and Tsunami with the height of 15 m on March 11, 2011. Small Modular Reactors (SMRs) have been developed to improve safety systems by utilizing passive...
Resource Type:: Dissertation
Full Text:: safety of nuclear power plants following the Fukushima Daiichi nuclear accident followed by 8.9

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

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

Scaling analysis of the OSU high temperature test facility during a pressurized conduction cooldown event using RELAP5-3D

Creator:: Castañeda, Juan A.
Abstract:: In early 2000, the Generation IV International Forum (GIF) was created to perform research and development for the next generation nuclear systems. Among the selected nuclear systems was the Very High Temperature Gas-Cooled Reactor (VHTR). Then in 2008, the U.S. Department of Energy (DOE) decided that the Next Generation Nuclear...
Resource Type:: Masters Thesis
Full Text:: started research and development into advanced nuclear reactor technologies. [1] Consequently, the

Proof of concept of the use of RAVEN and RELAP5-3D for risk informed safety margin characterization

Creator:: Riley, Thomas H.
Abstract:: This project was a proof of concept of the use of the RAVEN software, a tool developed for the Risk Informed Safety Margin Characterization (RISMC) approach, with RELAP5-3D. This novel approach combines older probabilistic and mechanistic approaches to look at how and why the complex systems of a nuclear power...
Resource Type:: Masters Thesis
Full Text:: probabilistic strength of safety margins and discovering and mapping paths to failure in Nuclear Power Plant

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

Cost-Benefit Analysis : Proposed Safety Upgrades to Currently Operating Nuclear Power Plants

Simulated antineutrino signatures of nuclear reactors for nonproliferation applications

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

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

Flow Dynamics and Condensation of Film Flows in Small Modular Reactors

RELAP5-3D modeling of ADS blowdown of MASLWR facility

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

Scaling analysis of the OSU high temperature test facility during a pressurized conduction cooldown event using RELAP5-3D

Proof of concept of the use of RAVEN and RELAP5-3D for risk informed safety margin characterization

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

Limit your search

Academic Affiliation

Advisor

Commencement Year

Committee Member

Creator

Date

Decade

Degree Field

Degree Level

Degree Name

File Format

Language

License

Non-Academic Affiliation

Peer Reviewed

Resource Type

Rights Statement

Subject

Search Constraints

Search Results

Limit your search