Doctoral Dissertations (FESNS)
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Browsing Doctoral Dissertations (FESNS) by Subject "Generation IV"
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Item A benchmarked dynamic model of xenon behavior in a molten salt reactor(2019-12-01) Price, Terry J.; Bereznai, George; Chvala, OndrejMolten salt reactors are a type of nuclear reactor that are being considered for deployment in the fourth generation nuclear power technology. Molten salt reactors use molten a alkali / actinide halide salt melt at temperatures far in excess of temperatures found in a typical pressurized water reactor. This thesis focuses on graphite moderated reactors with fluoride as the halide. The salt melt, called the fuel salt, is circulated between a moderator and a heat exchanger. While within the moderator, the dissolved actinides undergo fission and generate heat. Among products of nuclear fission is gaseous xenon, and in particular the isotope xenon-135 that acts as a neutron absorber. In solid fueled reactors, the xenon is effectively static and trapped within the fuel matrix. In a molten salt reactor, conversely, the fuel matrix is the mobile, circulating fuel salt that transports the xenon along with the rest of the fuel. This thesis focuses on modeling the behavior of xenon in a molten salt reactor. Existing literature in the field is reviewed and compiled. A model of xenon behavior in a molten salt reactor (the Molten Salt Reactor Experiment in particular) has been developed and the model is presented in this thesis. The model is benchmarked against experimental data using best available data, then minimal necessary justifiable adjustment is made to model parameters in order to fit the model to the experimental data. As a result this model is able to fit two transients, something that no xenon model of the molten salt reactor experiment has been able to do previously.Item Study on specifics of thermalhydraulics and neutronics of pressure-channel supercritical water-cooled reactors (SCWRs)(2017-08-01) Peiman, Wargha; Pioro, Igor; Gabriel, KamielA group of countries has initiated an international collaboration to develop a next generation (i.e., Generation IV) of nuclear reactors. Chosen as one of the six Generation‒IV nuclear-reactor concepts, the SCWRs are expected to have high thermal efficiencies within the range of 40 ‒ 50% owing to reactor’s high outlet temperatures. The Canadian pressure-tube-type SCWR is featured with 3-batch refueling, 336 vertical fuel channels, a porous ceramic insulator inside the pressure tube, and stainless-steel cladding. The reactor operates at a pressure of 25 MPa with the coolant temperature rising from 350 to 625°C. Consequently, sheath and fuel centerline temperatures are significantly higher in SCWRs compared to those of the current water-cooled nuclear reactors. The main objective of this thesis is to conduct a study on specifics of the thermalhydraulics and neutronics of a pressure-tube SCWR based on an understanding of the supercritical water phenomena and their impacts on reactor design and operation. This thesis investigates the impact of several thermalhydraulic modeling parameters on fuel and cladding temperatures of a pressure-tube SCWR. The investigated thermalhydraulic modeling parameters are: 1) variable heat transfer coefficient, which is affected by thermophysical properties of supercritical water, axial heat flux, and three heat-transfer regimes: normal, improved and deteriorated; 2) thermophysical properties, which are affected by the bulk-fluid-temperature profile along the heated length and pressure drop along the fuel channel; 3) variable axial and radial heat-flux profiles of a fuel assembly (bundle string), which are affected by the neutron flux; 4) radial non-uniform heat generation inside the fuel; 5) axial and radial variable thermal conductivity of a fuel; 6) contact thermal resistance between the fuel and cladding; 7) heat loss from the coolant to the moderator, which is affected by the thermal conductivity of a ceramic insert; and 8) pressure drop of the coolant along the fuel channel. The main neutronic aspects, which have been incorporated in the neutronic model, include 1) variable coolant density along the heated length of the fuel channel, which affects neutronic properties of a lattice and, hence, the neutron flux and 2) number of energy groups, which affects the calculated channel powers.