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dc.contributor.advisorPavel Hejzlar and Jacopo Buongiorno.en_US
dc.contributor.authorMemmott, Matthew Jen_US
dc.contributor.otherMassachusetts Institute of Technology. Dept. of Nuclear Science and Engineering.en_US
dc.date.accessioned2010-04-28T15:36:20Z
dc.date.available2010-04-28T15:36:20Z
dc.date.copyright2009en_US
dc.date.issued2009en_US
dc.identifier.urihttp://hdl.handle.net/1721.1/54463
dc.descriptionThesis (Ph. D.)--Massachusetts Institute of Technology, Dept. of Nuclear Science and Engineering, 2009.en_US
dc.descriptionCataloged from PDF version of thesis.en_US
dc.descriptionIncludes bibliographical references (p. 419-429).en_US
dc.description.abstractThe sodium fast reactor (SFR) is currently being reconsidered as an instrument for actinide management throughout the world, thanks in part to international programs such as the Generation-IV and especially the Global Nuclear Energy Partnership (GNEP). The success of these programs, in particular the GNEP, is dependent upon the ability of the SFR to manage actinide inventory while remaining economically competitive. In order to achieve these goals, the fuel must be able to operate reliably at high power densities. However, the power density of the fuel is limited by fuel-clad chemical interaction (FCCI) for metallic fuel, cladding thermal and irradiation strain, the fuel melting point, sodium boiling, and to a lesser extent the sodium pressure drop in the fuel channels. Therefore, innovative fuel configurations that reduce clad stresses, sodium pressure drops, and fuel/clad temperatures could be applied to the SFR core to directly improve the performance and economics. Two particular designs of interest that could potentially improve the performance of the SFR core are the internally and externally cooled annular fuel and the bottle-shaped fuel. In order to evaluate the thermal-hydraulic performance of these fuels, the capabilities of the RELAP5-3D code have been expanded to perform subchannel analysis in sodium-cooled fuel assemblies with non-conventional geometries. This expansion was enabled by the use of control variables in the code. When compared to the SUPERENERGY II code, the prediction of core outlet temperature agreed within 2%. In addition, the RELAP5-3D subchannel model was applied to the ORNL 19-pin test, and it was found that the code could predict the measured outlet temperature distribution with a maximum error of -8%.en_US
dc.description.abstract(cont.) As an application of this subchannel model, duct ribs were explored as a means of reducing core outlet temperature peaking within the fuel assemblies. The performance of the annular and bottle-shaped fuel was also investigated using this subchannel model. The annular fuel configurations are best suited for low conversion ratio cores. The magnitude of the power uprate enabled by metal annular fuel in the CR = 0.25 cores is 20%, and is limited by the FCCI constraint during a hypothetical flow blockage of the inner-annular channel due to the small diameters of the inner-annular flow channel (3.6 mm). On the other hand, a complete blockage of the hottest inner-annular flow channel in the oxide fuel case results in sodium boiling, which renders the annular oxide fuel concept unacceptable for use in a SFR. The bottle-shaped fuel configurations are best suited for high conversion ratio cores. In the CR = 0.71 cores, the bottle-shaped fuel configuration reduces the overall core pressure drop in the fuel channels by up to 36.3%. The corresponding increase in core height with bottle-shaped fuel is between 15.6% and 18.3%. A full-plant RELAP5-3D model was created to evaluate the transient performance of the base and innovative fuel configurations during station blackout and UTOP transients. The transient analysis confirmed the good thermal-hydraulic performance of the annular and bottle-shaped fuel designs with respect to their respective solid fuel pin cases.en_US
dc.description.statementofresponsibilityby Matthew J Memmott.en_US
dc.format.extent458 p.en_US
dc.language.isoengen_US
dc.publisherMassachusetts Institute of Technologyen_US
dc.rightsM.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.en_US
dc.rights.urihttp://dspace.mit.edu/handle/1721.1/7582en_US
dc.subjectNuclear Science and Engineering.en_US
dc.titleThermal-hydraulic analysis of innovative fuel configurations for the sodium fast reactoren_US
dc.typeThesisen_US
dc.description.degreePh.D.en_US
dc.contributor.departmentMassachusetts Institute of Technology. Dept. of Nuclear Science and Engineering.en_US
dc.contributor.departmentMassachusetts Institute of Technology. Department of Nuclear Science and Engineering
dc.identifier.oclc554672415en_US


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