Thesis (Ph. D.)--Lehigh University, 1996. / Includes vita. Bibliography: leaves 276-284.

Thesis (doctoral)--Rheinische Friedrich-Wilhelms-Universität Bonn, 1996. / Includes bibliographical references (p. 94-99).

Thesis--Wisconsin. / Vita. Includes bibliographical references (leaves 202-212).

No description available.

No description available.

No description available.

This research studies the effects of a postulated Main Steam Line Break accident in a nuclear plant on the transient loading of steam generator tubes. The problem involves complex transient two-phase flow dynamics and fluid-structural loading processes. A better understanding of this phenomenon will permit the development of improved design tools to ensure steam generator safety. This Thesis presents the results of an experimental laboratory study of the transient loading of a sectional model of steam generator tubes during a simulated two-phase blowdown. The research was carried out in a purpose designed and built experimental facility. The thermodynamic phenomena were investigated through dynamic pressures and temperatures. The transient tube loads were directly measured using dynamic load cells. The working fluid was R-134a and the tube bundle was a normal triangular array with a pitch ratio of 1.36. Preliminary testing indicated parasitic loading on the instruments and remedial actions were taken to ensure measurement accuracy. The success of the instrumentation development methodologies was validated in a series of single- phase blowdowns. Two-phase blowdown experiments were then conducted with various levels of liquid and numbers of tube rows. The results provided hitherto unknown information relating the nature of the fluid-structure interaction and flow development during a two-phase transient blowdown across a tube bundle. The pressure drop across the tube bundle established the thermal hydraulic fluid behaviour in the pressure vessel and controlled the upstream fluid discharge from the system. The flow through the bundle was choked for the majority of the transient. The transient tube loading was explained in terms of the associated fluid mechanics and the maximum load was compared with existing models obtained under steady flow conditions. An empirical model was developed that enables the prediction of the maximum tube loads once the pressure drop is known. / Thesis / Doctor of Philosophy (PhD) / The design of nuclear plant systems requires that public safety be ensured for the worst-case imaginable accident scenarios. This means that radioactive materials produced by nuclear fission must be safely contained. If the main steam pipe from a nuclear steam generator were to break, the water in the steam generator would rapidly boil off in what is called a blowdown. Such an event could produce significant loading on the heat exchanger tubes. Should the tubes rupture, radioactive materials may breach reactor containment. Thus, knowing the tube loading during this hypothetical scenario is an important input for safe design. The goal of this research is to improve our understanding of the effects of this postulated accident on the transient loading of the tubes. A laboratory study was carried out to simulate the blowdown. The experimental results provided valuable insights and guidance for the development of improved design tools.

No description available.

This paper presents the methods of predicting the steady-state two-phase flow (steam and water) pressure drop across the restrictions of annular geometries formed when tubes extend through circular holes in tube support plates. Two approaches are discussed and a detailed sample calculation of the one selected is presented. The major areas of discussion are the orientation of tubes-to-tube support plate holes, and the thickness of tube support plate. Finally, the conclusion gives a comparison of methods and recommendations for future investigations.

The flow patterns of an air-water mixture at atmospheric pressure were visually observed through a simulated vertical steam generator tube support plate of circular hole and trefoil designs. Flow oscillations with a period on the order of 0.4 seconds were observed in both support plates at similar mass flow rate combinations. The oscillation consisted of a period of rapid surging followed by a low flow or stagnation period on the order of 0.1 seconds. Reverse (downward) flow was also observed for part of the oscillation cycle in the trefoil support plate. The flow oscillations that exist in one tube support plate are coupled to the oscillation phase that exists in the preceding tube support plate. The operation of a steam generator in a region where this flow oscillation occurs could result in reduced to steam generator life due to increased corrosion.