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dc.contributor.authorLovett, Phyllis Mariaen_US
dc.contributor.authorTodreas, Neil E.en_US
dc.contributor.otherMassachusetts Institute of Technology. Department of Nuclear Engineeringen_US
dc.contributor.otherSandia National Laboratoriesen_US
dc.date.accessioned2014-09-16T23:38:01Z
dc.date.available2014-09-16T23:38:01Z
dc.date.issued1991en_US
dc.identifier.urihttp://hdl.handle.net/1721.1/89746
dc.description"September 1991."en_US
dc.descriptionAt head of title: Final report - experimentalen_US
dc.descriptionAlso issued as an M.S. thesis written by the first author and supervised by the second author, MIT Dept. of Nuclear Engineering, 1991en_US
dc.descriptionIncludes bibliographical references (pages 113-115)en_US
dc.description.abstractNuclear power reactors generate highly radioactive spent fuel assemblies. Initially, the spent fuel assemblies are stored for a period of several years in an on-site storage facility to allow the radioactivity levels of the assemblies to decay. As the radioactive fission product isotopes in the fuel decay, they generate significant amounts of thermal energy producing high temperatures in the spent fuel. The spent fuel from nuclear power plants will eventually have to be transferred to a federal geologic repository in a spent fuel transportation casks. The purpose of this research project is to characterize the relative importance of the heat transfer mechanisms of radiation, conduction, and convection in a dry horizontally-oriented nuclear spent fuel assembly, for eventual application in spent fuel transportation cask design.en_US
dc.description.abstractTo determine the relative importance of each heat transfer mode, an experiment was designed and operated to characterize the heat transfer in an 8x8 square heater rod array (similar to a Boiling Water Reactor fuel assembly) in a horizontal orientation. The experimental apparatus was operated with the following variable parameters and their ranges: Power to Heater Rods (Controlling Temperatures from 40'C to 250'C); Heater Transfer Medium (Air, Nitrogen, Argon, and Helium); Pressure of the Heat Transfer Medium (15 psig, 0 psig, 24 inches of mercury); Power to Boundary Condition Box (not controlled). The experiment was designed, fabricated, and operated under the Sandia National Laboratories-approved MIT Nuclear Engineering Department Quality Assurance Program developed in this work specifically for this project. The test data obtained from the experimental apparatus was analyzed with the lumped keff/hedge model developed by R.D.en_US
dc.description.abstractManteufel at MIT, in related work on this research project, and the Wooten-Epstein relationship developed at Battelle Memorial Institute. The test data was used to validate the lumped keff/hedge model. Good agreement was found between the lumped keff/hedge model and the test data in each Test Campaign with the exception of Below Atmospheric Pressure data. Both experimental and theoretical sources for the discrepancy are discussed. However, the full reason for the deviation is not know.en_US
dc.description.sponsorshipSponsored by Sandia National Laboratories contract # 42-5638en_US
dc.format.extent531 pagesen_US
dc.publisherCambridge, Mass. : Massachusetts Institute of Technology, Dept. of Nuclear Engineering, [1991]en_US
dc.relation.ispartofseriesMITNE ; no. 294en_US
dc.subject.lccTK9008.M41 N96 no.294en_US
dc.subject.lcshNuclear fuel elements -- Thermal propertiesen_US
dc.subject.lcshHeat -- Transmissionen_US
dc.subject.lcshSpent reactor fuelsen_US
dc.titleAn experiment to simulate the heat transfer properties of a dry, horizontal spent nuclear fuel assemblyen_US
dc.typeTechnical Reporten_US
dc.identifier.oclc857718061en_US


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