No description available.

This study has evaluated the degradation processes which will potentially affect the reinforced concrete structures of a nuclear power generation facility during and beyond its original design life. This task was undertaken in consideration of the feasibility of extending the life of nuclear power plants beyond their current license expiration dates. Following the identification of deterioration mechanisms which are expected to occur, an inspection system was developed to correctly assess and document the condition of the reinforced concrete components during their service life. Twenty-eight out of thirty-nine possible degradation modes are deemed likely to affect Surry’s reinforced concrete structures. The majority of these modes are visually evident in their incipient stages. Currently available nondestructive testing methods were assessed to determine their applicability to detect modes which are not visually evident or to determine the extent of deterioration due to other modes. It was found that many nondestructive testing methods are currently lacking in reliability, portability, or ease of application. Consequently, the developed inspection program is based primarily on visual inspections performed by qualified inspectors. This report was prepared under the authority of Virginia Power Company in conjunction with the Surry Unit One life extension study. It is the conclusion of this report that reinforced concrete degradation will in no way impair the usefulness or safety of the concrete structures of a nuclear facility during the 40 year design life provided actions are taken to implement a concrete inspection program similar to that which is described within. This program will allow the detection of potentially critical situations thereby directing the maintenance and repair activities necessary to insure the feasibility of extended life. / Master of Science

Thesis (M. Soc. Sc.)--University of Hong Kong, 1987.

Humboldt-Universiẗat, Diss.--Berlin, 2007. / Nebent.: Stilllegung von Kernkraftwerken. Includes bibliographical references (p. 247-264).

Thesis (M.Soc.Sc.)--University of Hong Kong, 1987. / Also available in print.

With the ever advancing state of computer systems, it is imperative to maintain the most up-to-date and reliable safety evaluation data for nuclear power systems. Commonplace now is the practice of updating old accident simulation results with more advanced models and codes using today's faster computer systems. Though it may be quite an undertaking, the benefits of using a more advanced model and code can be significant especially if the result of the new analysis provides increased safety margin for any plant component or system. A series of parametric and sensitivity studies for the Loss of Normal Feedwater Anticipated Transient without Scram (LONF ATWS) for Southern Company's Vogtle Electric Generating Plant (VEGP) Units 1&2 located near Waynesboro, GA was performed using the best-estimate thermal-hydraulics transient analysis code RETRAN-02w. This thesis includes comparison to the results of a generic plant study published by Westinghouse Electric Corporation in 1974 using an earlier code, LOFTRAN, as well as Vogtle-specific analysis. The comparative analysis exposes and seeks to explain differences between the two codes whereas the Vogtle analysis utilizes data from the Vogtle FSAR to generate plant-specific data. The purpose of this study is to validate and update the previous analysis and gather more information about the plant actions taken in response to a LONF ATWS. As a result, now there is a new and updated evaluation of the LONF ATWS for both a generic 4-loop Westinghouse plant and VEGP using a more advanced code. Beyond the reference case analysis, a series of sensitivity and parametric studies have been performed to show how well each type of plant is designed for handling an ATWS situation. These studies cover a wide range of operating conditions to demonstrate the dependability of the model. It was found that both the generic 4-loop Westinghouse PWR system and VEGP can successfully mitigate a LONF ATWS throughout the core's operating cycle.

Pressurizer modeling for predicting the dynamic pressure of the PIUS system is presented. The transient behavior of this model for the PIUS system was investigated. The validity of this model for the PIUS system is limited to transients that are neither too large nor too long in duration. For example, the model is not capable of describing events following a complete loss of liquid for the pressurizer. However, the model can be used for qualitative prediction of the PIUS system behavior for a wide variety of severe transients. A review of pressurizer modeling indicates that the neglecting of the change in the internal energy of the subcooled water during transients is an acceptable assumption. The inherently safe feature of the PIUS system was confirmed through the self-shutdown of the reactor or, in some cases, through reactor power reduction as a result of the ingress of the pool boric acid solution into the primary system. This dynamic model was constructed of three major components: (1) The primary loop, (2) The secondary loop, and (3) The natural convection loop through the pool. A lumped parameter model, uniform heat transfer, and point kinetics have been the main approximations in this model. Other approximations are mentioned during the modeling of each component of the model. The dynamic model was simulated using the DARE-P continuous system simulation language which was developed in the Electrical Engineering Department at the University of Arizona.

No description available.

The purpose of this research was a proof of concept using a fan motor stator as transducer to monitor motor rotor and attached axial fan for mechanical motion. The proof was to determine whether bearing faults and fan imbalances could be detected in vane-axial fans using Motor Electrical Signature Analysis (MESA). The data was statistically analyzed to determine if the MESA systems could distinguish between baseline conditions and discrete fault frequencies for the three test conditions: bearing inner race defect, bearing outer race defect, and fan imbalance. The statistical conclusions for these proofs of concept were that MESA could identify all three faulted conditions.

The vulnerability of electrical cables exposed to a fire environment is of particular concern to the nuclear power plant community. The community is interested in data that could be used for predicting cable failures during a fire situation. For this reason, a cable test program was conducted using two different types of cable insulation. Several different exposure heat fluxes were tested, as well as different test arrangements such as cable trays and conduits. The program revealed that a single failure temperature for all cable types is not recommended, but if it is necessary a reasonable temperature could be chosen for the thermosets tested in this project.