The phenomena of interest in this work is the thermal stratification which occurs during the early stages of a loss of coolant accident (LOCA) in the OSU APEX Thermal Hydraulic Test Facility, which is a scaled model of the Westinghouse AP600 nuclear power plant. Thermal stratification has been linked to the occurrence of pressurized thermal shock (PTS). Analysis of the OSU APEX facility data has allowed the determination of an onset criteria and support for the postulated mechanisms leading to thermal stratification. CFX 4.1, a computational fluid dynamics code, was used to generate a model of the cold legs and the downcomer and the phenomena occurring within them. The following are the accomplishments of the work contained within this report; Determined the causes of thermal stratification in the cold legs of the Westinghouse Advanced Passive 600 MW (AP600) nuclear power plant. Predicted the onset of thermal stratification in the cold legs of the Westinghouse Advanced Passive 600 MW (AP600) nuclear power plant. Modeled the phenomena associated with thermal stratification in the cold legs of the Westinghouse Advanced Passive 600 MW (AP600) nuclear power plant. / Graduation date: 1998

The Department of Nuclear Engineering at Oregon State University has performed a series of confirmatory tests for the United States Nuclear Regulatory Commission (USNRC). These tests have been conducted in the Advanced Plant Experiment (APEX) Facility which is a one quarter height scaled simulation of the Westinghouse Advanced Passive 600 megawatt electric (AP600) pressurized water reactor. The purpose of the testing program is to examine AP600 passive safety system performance, particularly during long term cooling. The NRC-25 series tests represents a parametric study to determine the minimum liquid reserves required to prevent a temperature excursion in the APEX core. The tests were initiated with a failure of all passive safety systems with the exception of portions of the fourth stage of the Automatic Depressurization System (ADS-4) and the In-containment Refueling Water Storage Tank (IRWST) injection system. These tests were concluded upon the onset of IRWST injection or core uncovery as determined by a temperature excursion in the core. The purpose of this thesis is to present the results of the NRC-25 test series. This includes a theoretical model which was developed to predict core liquid inventory during an ADS-4 blowdown. The NRC-25 test series was used to benchmark a theoretical model derived using the mass and energy conservation equations, the perfect gas law, and a critical gas flow model. It will be shown that this model agrees with the experimental data quite well. / Graduation date: 1999

Thesis (Ph. D.)--University of Missouri-Columbia, 2002. / Typescript. Vita. Includes bibliographical references (leaves 122-128). Also available on the Internet.

Thesis (Ph. D.)--University of Missouri-Columbia, 2002. / Typescript. Vita. Includes bibliographical references (leaves 122-128). Also available on the Internet.

An on-line, passive method for monitoring the moderator temperature coefficient (MTC) in a pressurized water reactor is described and compared to an analytical model. The method correlates fluctuations in the neutron flux with fluctuations in the coolant temperature. Results show that the frequency response function between the in-core neutron flux and thermocouple signals is proportional to the MTC and can be used to monitor changes in the MTC throughout the fuel cycle. An analytical model is also formulated and used to investigate how the neutron flux and coolant temperature fluctuations are related and change in the axial core direction. The analytical model results agree reasonably well with the experimental results. / Ph. D.

A computer program has been developed which performs a thermal-hydraulic analysis of a mechanical spectral shift reactor fuel assembly. In this reactor type, the fuel to water volume ratio is changed to shift the neutron energy spectrum and control reactivity. The ability to handle different fuel to water ratios is the unique feature of this program. The required input parameters are the coolant inlet pressure and temperature, the mass flow rates, and the power density. The cladding temperature and the coolant pressure and temperature are calculated throughout the fuel assembly. Three test cases were performed using a conceptual reactor core design. The fuel to water ratio of this design can be varied from 1.25 to 0.75 by removing blank rods from the core. Each fuel assembly is enclosed in a hex can. The fuel assembly flowrates can be independently selected. The test case results indicate that proper flowrate selection can produce equal fuel assembly enthalpy gains. The maximum allowable pressure drop determines the maximum coolant flowrate and the maximum power density variation between fuel assemblies. The coolant exit properties are relatively unaffected by changing the fuel to water ratio. This indicates that little heat is transferred into the moderator regions that are formed when void rods are removed. The most useful program modification would be the addition of a routine that defines the coolant channel parameters for any fuel assembly. / Master of Science

Under simulated accident conditions, tees in the primary coolant loop of a Pressurized Water Reactor (PWR) can deviate from their original design purpose and become separators that effectively remove core heat sink capacity. This method of primary coolant removal is a phenomelogical subset of phase separation known as liquid entrainment, whereby liquid is forced from its original path by the inertia of the gas. A comprehensive literature review revealed common deficiencies in previous studies. The Westinghouse AP600 advanced reactor design was chosen to assess the validity of entrainment models. Following a systematic scaling analysis of the prototypic design a model separate effects test was proposed and constructed at Oregon State University. Just under 100 tests were run to fill the deficiencies found in the literature review. New data from the Air-water Test Loop for Advanced Thermal-hydraulic Studies (ATLATS) could not be predicted by published correlations. A new theoretical model for predicting liquid entrainment onset and steady state entrainment was developed. Comparison with all available data shows a marked improvement for predicting the mass flow rate out the vertical branch. / Graduation date: 2003

Critical Heat Flux (CHF) is an important safety parameter for the design of nuclear reactors. The most commonly used predictive tool for determination of CHF is a look-up table developed using tube data with an average hydraulic test diameter of 8 mm. There exist in the world today nuclear reactors whose geometry is annular, not tubular, and whose hydraulic diameter is significantly smaller than 8 mm. In addition, any sub-channel thermal hydraulic model of fuel assemblies is annular and not tubular. Comparisons were made between this predictive tool and annular correlations developed from test data. These comparisons showed the look-up table over-predicts the CHF values for annular channels, thus questioning its ability to perform correct safety evaluations. Since no better tool exists to predict CHF for annular geometry, an effort was undertaken to produce one. A database of open literature annular CHF values was created as a basis for this new tool. By compiling information from eighteen sources and requiring that the data be inner wall, unilaterally, uniformly heated with no spacers or heat transfer enhancement devices, a database of 1630 experimental values was produced. After a review of the data in the database, a new look-up table was created. A look-up table provides localized control of the prediction to overcome sparseness of data. Using Shepard's Method as the extrapolation technique, a regular mesh look-up table was produced using four main variables: pressure, quality, mass flux, and hydraulic diameter. The root mean square error of this look-up table was found to be 0.8267. However, by fixing the hydraulic diameter locations to the database values, the root mean square error was further reduced to 0.2816. This look-up table can now predict CHF values for annular channels over a wide range of fluid conditions. / Graduation date: 2001

Steam interfacial condensation in a core makeup tank was simulated using the code RELAP5/MOD3 version 8.0 to predict the violent pressure oscillation phenomena in a core makeup tank. Six base cases were carried out to study the effects of back pressure and of vacuum conditions produced in the core makeup tank by rapid steam condensation. The effect of varying the liquid conduction thermal layer thickness was studied. In addition, the code's ability to predict condensation heat transfer was evaluated. Violent pressure oscillations were found in the early period of a transient. The violent pressure oscillations had no effect on the total amount of injection water from core makeup tank. The conduction thermal layer thickness was found to only effect the liquid temperature history. The current version of RELAP5/MOD3 was found to be incapable of dealing with the condensation heat transfer problem in which the volume liquid temperature is lower than the temperature of the heat structure which is connected to that hydraulic volume. / Graduation date: 1994

published_or_final_version / Mechanical Engineering / Master / Master of Philosophy