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dc.contributor.advisorAl-Sheikhly, Mohamaden_US
dc.contributor.advisorSilverman, Josephen_US
dc.contributor.authorMohamed, Ali Bellouen_US
dc.date.accessioned2010-02-19T07:09:36Z
dc.date.available2010-02-19T07:09:36Z
dc.date.issued2009en_US
dc.identifier.urihttp://hdl.handle.net/1903/10007
dc.description.abstractTwo MCNP models (detailed and approximated) of University of Maryland Training Reactor were created. The detailed model attempted to simulate the reactor according to engineering specifications while the simplified model eliminated all structural materials above and below the core. Neutron flux spectrum calculations for both models within the core showed that the results obtained from both models agreed within less than 0.5%. It was concluded that reactors equipped with standard TRIGA fuels enriched to 20 percent in uranium-235 can be modeled with all structures above and below the core eliminated entirely from the model without increasing the error due to geometry modeling simplifications of the core. In TRIGA reactors supplied with standard TRIGA fuels enriched to 20 percent in U-235, the graphite reflectors above and below the fuel act as "neutron energy regulators." Neutrons reflected back into the core through the graphite reflectors quickly become thermalized even if their energies were altered due to the change in materials properties above and below the core. Both MCNP models results agree well with measured data. It was also found that simplification in the target geometry leads to substantial uncertainty in the calculated results. The neutron energy spectrum, thermal flux, and total flux were calculated at the thermal column access plug face; in the pneumatic transfer system rabbit, and on top and bottom sections of the most center fuel element. The thermal flux and the total flux at the thermal column access plug face both agreed with measured data within a 5% uncertainty. The thermal flux, fast flux, and the total flux in the rabbit differ by 18.8%, 35%, and 5.7% respectively, from the measured data. The relatively high uncertainty (in the neutron energy distribution but not the total neutron flux) was attributed to the use of air as the target irradiated inside the rabbit. For such a thin target (15 mg/cm2), a precise neutron balance between reflection and absorption events is difficult to obtain; that will alter the thermal or fast flux values. The contribution of this work to the reactor users is that a virtual reactor model that compared well with experiment is created. Experiments utilizing the reactor experimental facilities (thermal column, through tube, pneumatic transfer system rabbit, and beam ports) can now be optimized before they are executed. The contribution of this work to the research reactor community at is that research reactors equipped with standard TRIGA fuels can be modeled with core geometry approximations, such as these adopted in this work, without affecting the precision and accuracy of the Monte Carlo calculations.en_US
dc.titleFlux Maps Obtained from Core Geometry Approximations: Monte Carlo Simulations and Benchmark Measurements for a 250 kW TRIGA Reactoren_US
dc.typeDissertationen_US
dc.contributor.publisherDigital Repository at the University of Marylanden_US
dc.contributor.publisherUniversity of Maryland (College Park, Md.)en_US
dc.contributor.departmentMaterial Science and Engineeringen_US
dc.subject.pqcontrolledEngineering, Nuclearen_US


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