The OSU/APEX thermal hydraulic test facility models the passive safety systems of the Westinghouse AP600 advanced light water reactor design. Numerous experiments have been performed to test these systems, the one of focus here is the station blackout scenario. This experiment simulated the complete loss of AC power to all plant systems. One of the objectives of this experiment was to determine the effectiveness of the Passive Residual Heat Removal (PRHR) system. The PRHR system removes heat by rejecting it into the In-containment Refueling Water Storage Tank (IRWST). The IRWST houses the PRHR and is used as a heat sink for the decay heat. The PRHR is a C-type tube heat exchanger. Heat is removed through two mechanisms: natural convection and nucleate boiling from the surface of the PRHR. As the experiment progressed, a large degree of thermal stratification was observed in the IRWST with no significant thermal mixing. A thermal layer developed in the top of the tank and as the thermal layer approached saturation the rate of heat removal from the sections of the PRHR engulfed by this layer decreased. The effectiveness of these sections of the PRHR continued to decrease until unexpected flow patterns developed at the same time that the thermal layer reached saturation. The IRWST fluid exhibited a bulk azimuthal flow pattern that increased the effectiveness of the PRHR. This increase allowed for more heat to be injected into the IRWST. However, the bulk fluid motion still did not mix the thermal layers. A three-dimensional computational fluid dynamic model using the CFX-4.2 software was developed to study the PRHR/IRWST system. The model uses the RPI method to account for the sub-cooled boiling that is present on the PRHR surface. The model successfully predicted the thermal stratification in the IRWST to within 4 K of experimental data. A counter-current flow was shown to occur along the interface of the thermal layers. This caused an enhancement of the heat transfer and turbulent mixing occurring across the interface of the thermal layers. / Graduation date: 1998

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