Thesis (M.S.)--University of Wisconsin--Madison, 1981. / Typescript. eContent provider-neutral record in process. Description based on print version record. Includes bibliographical references (leaves 159-163).

Measurements of mean velocity and mean temperature fields and wall parameters for air flowing in a smooth, vertical tube at low entry Reynolds numbers are presented for heating with constant wall heat flux along the heated length. Two entry Reynolds numbers of approximately 6000 and 4000 were employed with three heating rates, q('+) = q('w'')/ (Gc(,p,i) T(,i)), of approximately 0.0018, 0.0035 and 0.0045. The flow development was measured by obtaining internal profiles along the heated length at axial locations from x/D = 3.17 to x/D = 24.54. An adiabatic entry of 50 diameters preceded the heated region. The three heating rates caused slight, large and severe property variation of the air. The highest heating rate was found to cause significant buoyancy effects. The internal measurements were obtained using constant temperature hot-wire anemometry and resistance thermometry for velocity and temperature, respectively, employing a single short wire probe. The technique developed and employed for the use of a single short hot wire in velocity measurements in non-isothermal flows is presented. The measurements are compared to numerical predictions employing two simple versions of the van Driest mixing length turbulence model. In general, both models agreed with the measurements reasonably well, but for the higher heating rates neither model was completely satisfactory in predicting the velocity profiles. When the buoyancy parameter reached 0.3, the peak velocity occurred in the wall region rather than at the tube centerline. Typically, the Nusselt number was overpredicted by 10% for x/D > 14 and, consequently, the wall temperature was underpredicted by about 7%.

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Thesis (doctoral)--Technical University of Delft, 2007. / "Gas-cooled fluidized bed nuclear reactor ... (FLUBER)"--P. 3. Includes bibliographical references (p. [139]-148).

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Since the 1990s, researchers around the world have been creating antineutrino detectors for monitoring power reactors. These detectors have been deployed at light water reactors and are able to determine power levels and burn up throughout a fuel cycle. This technology could allow the IAEA to monitor LWRs remotely and unobtrusively to determine if they are operating using normal parameters. Very soon, the next generation of detector will be deployed at a CANDU reactor for a trial operation. While physical observation of these detectors is necessaryl in determining their usefulness, reactor physics simulations have proven to be very accurate in their prediction of detector performance. Since there are many designs still in development, reactor physics simulations are the only way to determine the efficacy of the detector technology. In addition to this, reactor simulations are the best way to evaluate the detector technology to ascertain its usefulness during diversion scenarios. In this research, antineutrino source terms were calculated for a High Temperature Gas Cooled Reactor core. These source terms were a function of power level and initial enrichment. SCALE6.1, developed by Oak Ridge National Laboratory, was used to calculate the isotopic inventory in the core as a function of depletion. These fertile and fissile isotopics, along with the fission cross-section and number of antineutrinos emitted per fission, were used to predict the antineutrino source rate for the core. It was found that changing the power yields a linear response from the antineutrino source term. By increasing the power by five percent, the source term also increased by five percent. Substantial changes in the initial enrichment also lead to a detectable change in the antineutrino source term. / Graduation date: 2012

Among the most formidable challenges facing our world is the need for safe, clean, affordable energy sources. Growing concerns over global warming induced climate change and the rising costs of fossil fuels threaten conventional means of electricity production and are driving the current nuclear renaissance. One concept at the forefront of international development efforts is the High Temperature Gas-Cooled Reactor (HTGR). With numerous passive safety features and a meltdown-proof design capable of attaining high thermodynamic efficiencies for electricity generation as well as high temperatures useful for the burgeoning hydrogen economy, the HTGR is an extremely promising technology. Unfortunately, the fundamental understanding of neutron behavior within HTGR fuels lags far behind that of more conventional watercooled reactors. HTGRs utilize a unique heterogeneous fuel element design consisting of thousands of tiny fissile fuel kernels randomly mixed with a non-fissile graphite matrix. Monte Carlo neutron transport simulations of the HTGR fuel element geometry in its full complexity are infeasible and this has motivated the development of more approximate computational techniques. A series of MATLAB codes was written to perform Monte Carlo simulations within HTGR fuel pebbles to establish a comprehensive understanding of the parameters under which the accuracy of the approximate techniques diminishes. This research identified the accuracy of the chord length sampling method to be a function of the matrix scattering optical thickness, the kernel optical thickness, and the kernel packing density. Two new Monte Carlo methods designed to focus the computational effort upon the parameter conditions shown to contribute most strongly to the overall computational error were implemented and evaluated. An extended memory chord length sampling routine that recalls a neutron’s prior material traversals was demonstrated to be effective in fixed source calculations containing densely packed, optically thick kernels. A hybrid continuous energy Monte Carlo algorithm that combines homogeneous and explicit geometry models according to the energy dependent optical thickness was also developed. This resonance switch approach exhibited a remarkably high degree of accuracy in performing criticality calculations. The versatility of this hybrid modeling approach makes it an attractive acceleration strategy for a vast array of Monte Carlo radiation transport applications. / text

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Thesis (Ph. D.)--University of Texas at Austin, 2008. / Vita. Includes bibliographical references.