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Lin-wen Hu and Mujid S. Kazimi. Wong, Susanna Yuen-Ting Massachusetts Institute of Technology. Dept. of Nuclear Science and Engineering. Massachusetts Institute of Technology. Department of Nuclear Science and Engineering 2010-03-25T15:24:30Z 2010-03-25T15:24:30Z 2008 2008 http://hdl.handle.net/1721.1/53275 547275119 Thesis (S.M. and S.B.)--Massachusetts Institute of Technology, Dept. of Nuclear Science and Engineering, 2008. Page no. "2" used twice. Cataloged from PDF version of thesis. Includes bibliographical references (p. 147-150 [i.e. p. 148-151]). The MIT Nuclear Research Reactor (MITR) is the only research reactor in the United States that utilizes plate-type fuel elements with longitudinal fins to augment heat transfer. Recent studies on the conversion to low-enriched uranium (LEU) fuel at the MITR, together with the supporting thermal hydraulic analyses, propose different fuel element designs for optimization of thermal hydraulic performance of the LEU core. Since proposed fuel design has a smaller coolant channel height than the existing HEU fuel, the friction pressure drop is required to be verified experimentally. The objectives of this study are to measure the friction coefficient in both laminar and turbulent flow regions, and to develop empirical correlations for the finned rectangular coolant channels for the safety analysis of the MITR. A friction pressure drop experiment is set-up at the MIT Nuclear Reactor Laboratory, where static differential pressure is measured for both flat and finned coolant channels of various channel heights. Experiment data show that the Darcy friction factors for laminar flow in finned rectangular channels are in good agreement with the existing correlation if a pseudo-smooth equivalent hydraulic diameter is considered; whereas a new friction factor correlation is proposed for the friction factors for turbulent flow. Additionally, a model is developed to calculate the primary flow distribution in the reactor core for transitional core configuration with various combinations of HEU and LEU fuel elements. by Susanna Yuen-Ting Wong. S.M. S.B. 150 [i.e. 151] p. eng Massachusetts Institute of Technology M.I.T. theses are protected by copyright. They may be viewed from this source for any purpose, but reproduction or distribution in any format is prohibited without written permission. See provided URL for inquiries about permission. http://dspace.mit.edu/handle/1721.1/7582 Nuclear Science and Engineering. Friction pressure drop measurements and flow distribution analysis for LEU conversion study of MIT Research Reactor Thesis application/pdf false Publication unknown unknown <Publication xmlns="https://www.openaire.eu/cerif-profile/1.1/" id="58d1d8a5-732a-408e-85dd-fb822afe0279"> <Type xmlns="https://www.openaire.eu/cerif-profile/vocab/COAR_Publication_Types">http://purl.org/coar/resource_type/c_1843</Type> <Language>eng</Language> <Title>Friction pressure drop measurements and flow distribution analysis for LEU conversion study of MIT Research Reactor</Title> <PublishedIn> <Publication> </Publication> </PublishedIn> <PublicationDate>2008</PublicationDate> <Authors> <Author> <DisplayName>Wong, Susanna Yuen-Ting</DisplayName> <Affiliation> <OrgUnit> </OrgUnit> </Affiliation> </Author> </Authors> <Editors> </Editors> <Publishers> <Publisher> <DisplayName>Massachusetts Institute of Technology</DisplayName> <OrgUnit /> </Publisher> </Publishers> <License>http://dspace.mit.edu/handle/1721.1/7582</License> <Keyword>Nuclear Science and Engineering.</Keyword> <Abstract>The MIT Nuclear Research Reactor (MITR) is the only research reactor in the United States that utilizes plate-type fuel elements with longitudinal fins to augment heat transfer. Recent studies on the conversion to low-enriched uranium (LEU) fuel at the MITR, together with the supporting thermal hydraulic analyses, propose different fuel element designs for optimization of thermal hydraulic performance of the LEU core. Since proposed fuel design has a smaller coolant channel height than the existing HEU fuel, the friction pressure drop is required to be verified experimentally. The objectives of this study are to measure the friction coefficient in both laminar and turbulent flow regions, and to develop empirical correlations for the finned rectangular coolant channels for the safety analysis of the MITR. A friction pressure drop experiment is set-up at the MIT Nuclear Reactor Laboratory, where static differential pressure is measured for both flat and finned coolant channels of various channel heights. Experiment data show that the Darcy friction factors for laminar flow in finned rectangular channels are in good agreement with the existing correlation if a pseudo-smooth equivalent hydraulic diameter is considered; whereas a new friction factor correlation is proposed for the friction factors for turbulent flow. Additionally, a model is developed to calculate the primary flow distribution in the reactor core for transitional core configuration with various combinations of HEU and LEU fuel elements.</Abstract> <Access xmlns="http://purl.org/coar/access_right" > </Access> </Publication>