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1	A heterogeneous finite element method and a leakage corrected homogenization technique Nichita, Eleodor Marian 12 1900 (has links) No description available. Light water reactors Safety measures Nuclear power plants Safety measures
2	RELAP5-3D modeling of ADS blowdown of MASLWR facility Bowser, Christopher Jordan 13 June 2012 (has links) 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 depressurization system (ADS) lines connect the primary vessel to a high pressure containment (HPC) which dissipates steam heat through a heat transfer plate thermally connected to another vessel with a large cool water inventory. This feature drew the interest of the IAEA and an ICSP was developed where a loss of feedwater to the steam generators prompted a depressurization of the primary vessel via a blowdown through the ADS lines. The purpose of the ICSP is to evaluate the applicability of thermal-hydraulic computer codes to unique experiments usually outside of the validation matrix of the code itself. RELAP5-3D 2:4:2 was chosen to model the ICSP. RELAP5-3D is a best-estimate code designed to simulate transient fluid and thermal behavior in light water reactors. Modeling was conducted in RELAP5-3D to identify the strengths and weaknesses of the code in predicting the experimental trends of the IAEA ICSP. This extended to nodalization sensitivity studies, an investigation of built-in models and heat transfer boundary conditions. Besides a qualitative analysis, a quantitative analysis method was also performed. / Graduation date: 2013 Nuclear Thermal Light water reactors -- Safety measures Pressurized water reactors -- Blowdown Nuclear reactors -- Containment
3	High pressure condensation heat transfer in the evacuated containment of a small modular reactor Casey, Jason R. 19 December 2012 (has links) 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. The design includes an initially evacuated, high pressure capable containment system simulated by a 5 meter tall pressure vessel. The convection-condensation process that occurs during use of the Emergency Core Cooling System has been characterized during two experimental continuous blowdown events. Experimental data has been used to calculate an average heat transfer coefficient for the containment system. The capability of the containment system has been analytically proven to be a conservative estimate of the full scale reactor system. / Graduation date: 2013 Condensation Nuclear Small Modular Reactor Containment Light water reactors -- Safety measures Nuclear reactors -- Containment Condensation Heat -- Transmission
4	Feed-and-bleed transient analysis of OSU APEX facility using the modern Code Scaling, Applicability, and Uncertainty method Hallee, Brian Todd 05 March 2013 (has links) 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, Applicability, and Uncertainty (CSAU) methodology has been applied to more accurately predict the safety margins of the Oregon State University Advanced Plant Experiment (APEX) facility experiencing a Loss-of-Feedwater Accident (LOFA). The statistical advantages of the Bayesian paradigm of probability was utilized to incorporate prior knowledge when determining the analysis required to justify the safety margins. RELAP5 Mod 3.3 was used to accurately predict the thermal-hydraulics of a primary Feed-and-Bleed response to the accident using assumptions to accompany the lumped-parameter calculation approach. A novel coupling of thermal-hydraulic and statistical software was accomplished using the Symbolic Nuclear Analysis Package (SNAP). Uncertainty in Peak Cladding Temperature (PCT) was calculated at the 95/95 probability/confidence levels under a series of four separate sensitivity studies. / Graduation date: 2013 APEX BEPU CSAU DAKOTA EMDAP feed and bleed LOCA integral test facility integral effects test light water reactor pressurized water reactor PCT peak cladding temperature RELAP response surface bayesian statistics wilks method formula NRC uncertainty quantification sensitivity analysis thermal hydraulic safety analysis Light water reactors -- Risk assessment

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