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  • About
  • The Global ETD Search service is a free service for researchers to find electronic theses and dissertations. This service is provided by the Networked Digital Library of Theses and Dissertations.
    Our metadata is collected from universities around the world. If you manage a university/consortium/country archive and want to be added, details can be found on the NDLTD website.
1

Ratchetting strains in pressurised pipes

Charles, Ian David January 1992 (has links)
No description available.
2

An assessment of thermal hydraulic analysis methods for pressurized thermal shock evaluations

Young, Eric P. 01 May 2002 (has links)
Improved methods of determining temperature transients in reactor systems are desired because of recent interest in Pressurized Thermal Shock (PTS) issues. The research presented herein was performed in support of the Nuclear Regulatory Commission's effort to re-evaluate its existing PTS rules. These rules are particularly important to the re-licensing of aging nuclear power plants. The much advanced computational power available to industry may offer a tool that allows the accurate calculation of temperatures inside the reactor vessel while not being inaccessibly expensive. It is proposed that an off-the-shelf Computational Fluid Dynamic (CFD) code, STAR-CD, can be a competitive tool in solving the thermal hydraulic domain of a reactor system. A comparison of the methodology and accuracy of the code types that have been previously used in PTS and one that has not been used extensively, CFD, is provided. A review of the literature shows that computer codes have been validated for solving PTS scenarios. The highly specialized program, REMIX, has been utilized extensively from 1986 to 1991 to interpret accident scenarios in reactor systems. Other programs are also available that can calculate downcomer temperatures including system and CFD type codes. Three codes representing the three different types of programs available are described in detail in the literature review section. Data appropriate for assessing a program's ability to calculate the response of a system to a PTS scenario is available from the current matrix of PTS tests being completed at the APEX-CE facility of the Oregon State University Nuclear Engineering department. The facility is a reduced scale integral test facility originally built for modeling the then-proposed AP-600 plant designed by Westinghouse. For the current test series, the facility was modified to model the Palisades nuclear power plant, a Combustion Engineering Pressurized Water Reactor (PWR). Two of the tests were chosen for their PTS typical conditions to compare with calculations of STAR-CD, REMIX, and RELAP. The computer models in each of the programs were either created, modified from a previous version, or the calculations for the comparisons were contributed. The downcomer temperatures at several locations and cold leg temperature gradients, where available, were extracted from the data and calculations and compared. Comparisons are presented in chapter 5 with graphs, along with some interpretation of the comparisons. It was found that STAR-CD agreed best with the data set in the downcomer and is the only program that calculated the temperature gradient in the cold legs. The agreement of STAR-CD with the cold leg data is also very good. REMIX and RELAP calculations agreement with data for downcomer temperatures are found to be good for all comparisons made, qualitatively more than quantitatively when contrasted with the STAR-CD calculations. / Graduation date: 2002
3

Anisotropic mechanical behaviour of a Zr-Sn-Nb-Mo alloy

Salinas Rodríguez, Armando January 1984 (has links)
No description available.
4

Anisotropic mechanical behaviour of a Zr-Sn-Nb-Mo alloy

Salinas Rodríguez, Armando January 1984 (has links)
No description available.
5

Top-down scaling analysis of the integral reactor vessel test facility

Graves, Joshua D. 07 December 2012 (has links)
Oregon State University has conducted research in collaboration with TerraPower, LLC, to perform a top-down scaling analysis of an integrated test facility. The goal of this facility is to simulate transient and quasi-steady phenomena at a reduced scale, including steady-state operation, pump coastdown, natural circulation, reactor head heat transfer, and coolant stratification. To support this goal, this thesis presents the methodology and analysis by which approximate facility dimensions were generated. This analysis includes implementation of the hierarchical two-tiered scaling methodology, as outlined by the Nuclear Regulatory Commission and optimization through the general reduced gradient methodology. / Graduation date: 2013
6

A discussion of the problems involved in the defection [sic] of fuel element leaks in a sodium cooled reactor [Masters thesis] /

Wegst, Walter Frederick, January 1957 (has links)
Thesis (M.S.)--University of Michigan, 1957.
7

Improvements to the pool critical assembly benchmark using 3-D discrete ordinate transport with adaptive difference

Edgar, Christopher Austin 20 September 2013 (has links)
The internationally circulated Pool Critical Assembly (PCA) Pressure Vessel Benchmark was analyzed using the PENTRAN Parallel SN code system for the geometry, material, and source specifications as described in the PCA Benchmark documentation. Improvements to the benchmark are proposed through the application of more representative flux and volume weighted homogenized cross sections for the PCA reactor core, which were obtained from a rigorous heterogeneous modeling of all fuel assembly types in the core. A new source term definition is also proposed based on calculated relative power in each core fuel assembly with a spectrum based on the Uranium-235 fission spectra. This research focused on utilizing the BUGLE-96 cross section library and accompanying reaction rates, while examining both adaptive differencing on a coarse mesh basis, as well as the sole use of Directional Theta-Weighted (DTW) SN differencing scheme in order to compare the calculated PENTRAN results to measured data. The results show good comparison with the measured data, which suggests PENTRAN is a viable and reliable code system for calculation of light water reactor neutron shielding and dosimetry calculations. Furthermore, the improvements to the benchmark methodology resulting from this work provide a 6 percent increase in accuracy of the calculation (based on the average of all calculation points), when compared with experimentally measured results at the same spatial location in the PCA pressure vessel simulator.
8

Nonlinear ultrasound for radiation damage detection

Matlack, Kathryn H. 01 April 2014 (has links)
Radiation damage occurs in reactor pressure vessel (RPV) steel, causing microstructural changes such as point defect clusters, interstitial loops, vacancy-solute clusters, and precipitates, that cause material embrittlement. Radiation damage is a crucial concern in the nuclear industry since many nuclear plants throughout the US are entering the first period of life extension and older plants are currently undergoing assessment of technical basis to operate beyond 60 years. The result of extended operation is that the RPV and other components will be exposed to higher levels of neutron radiation than they were originally designed to withstand. There is currently no nondestructive evaluation technique that can unambiguously assess the amount of radiation damage in RPV steels. Nonlinear ultrasound (NLU) is a nondestructive evaluation technique that is sensitive to microstructural features such as dislocations, precipitates, and their interactions in metallic materials. The physical effect monitored by NLU is the generation of higher harmonic frequencies in an initially monochromatic ultrasonic wave, arising from the interaction of the ultrasonic wave with microstructural features. This effect is quantified with the measurable acoustic nonlinearity parameter, beta. In this work, nonlinear ultrasound is used to characterize radiation damage in reactor pressure vessel steels over a range of fluence levels, irradiation temperatures, and material composition. Experimental results are presented and interpreted with newly developed analytical models that combine different irradiation-induced microstructural contributions to the acoustic nonlinearity parameter.
9

A performance model for a helically coiled once-through steam generator tube

Bayless, Paul David January 1979 (has links)
Thesis (M.S.)--Massachusetts Institute of Technology, Dept. of Nuclear Engineering, 1979. / This electronic version was submitted by the student author. The certified thesis is available in the Institute Archives and Special Collections. / MICROFICHE COPY AVAILABLE IN ARCHIVES AND SCIENCE. / Includes bibliographical references. / by Paul David Bayless. / M.S.

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