The purpose of this paper is to describe the flow oscillations which occur in the AP600 long term cooling test facility at Oregon State University. The AP600 system is an advanced pressurized water reactor design utilizing passive emergency cooling systems. A few hours after the initiation of a cold leg break, the passive cooling systems inject gravity fed cold water at a rate allowing steam production in the reactor vessel. Steam production in the core causes the pressure in the upper head to increase leading to flow oscillations in all the connecting reactor systems. This paper will show that the oscillations have a definite region of onset and termination for specific conditions in the APEX testing facility. Tests performed at high powers, high elevation breaks, and small break sizes do not exhibit oscillations. The APOS (Advanced Plant Oscillation Simulator) computer code has been developed using a quasi-steady state analysis for flows and a transient analysis for the core node energy balance. The pressure in the reactor head is calculated using a modified perfect gas analysis. For tank liquid inventories, a simple conservation of mass analysis is used to estimate the tank elevations. Simulation logic gleaned from APEX data and photographic evidence have been incorporated into the code to predict termination of the oscillations. Areas which would make the work more complete include a better understanding of two-phase fluid behavior for a top offtake on a pipe, more instrumentation in the core region of the APEX testing facility, and a clearer understanding of fluid conditions in the reactor barrel. Scaling of the oscillations onset and pressure amplitude are relatively straight forward, but termination and period are difficult to scale to the full AP600 plant. Differences in the core power profile and other geometrical differences between the testing facility and the actual plant make the scaling of this phenomenon to the actual plant conditions very difficult. / Graduation date: 1997

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 plant conditions while in the shutdown mode. In particular, to determine the time to "core uncovery" without operator interaction. The other objective is to study the effect of RHR heat exchanger elevation on natural circulation mass flow rate, fluid temperature and peak fuel pin temperature. Peak temperature and time to core uncovery were found for two small break LOCA cases. The natural circulation mass flow rate after break initiation was affected by varying the RHR heat exchanger elevation. The system pressure and temperature were not affected much by the elevation change in the RHR heat exchanger. The current version of RELAP5/MOD3 was found to be sensitive to the initial conditions in studies of low pressure,low temperature plant systems, especially for a large break LOCA. / Graduation date: 1994

published_or_final_version / Public Administration / Master / Master of Social Sciences

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Thesis (B.S.)--University of Missouri, School of Mines and Metallurgy, 1913. / The entire thesis text is included in file. Typescript. Illustrated by authors. Title from title screen of thesis/dissertation PDF file (viewed March 31, 2009) Includes bibliographical references (p. 22).

Thesis (M.S.)--University of Missouri-Columbia, 2008. / The entire dissertation/thesis text is included in the research.pdf file; the official abstract appears in the short.pdf file (which also appears in the research.pdf); a non-technical general description, or public abstract, appears in the public.pdf file. Title from title screen of research.pdf file (viewed on September 12, 2008) Includes bibliographical references.

Thesis (M.S.)--University of Wisconsin--Madison, 1955. / Published in Denver by the Technical Information Branch, Denver Federal Center, 1954. eContent provider-neutral record in process. Description based on print version record.

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