Nuclear Engineering - Master's degree
http://hdl.handle.net/1721.1/7689
2015-04-28T06:15:47ZCFD in support of development and optimization of the MIT LEU fuel element design
http://hdl.handle.net/1721.1/95601
CFD in support of development and optimization of the MIT LEU fuel element design
Diaconeasa, Mihai Aurelian
The effect of lateral power distribution of the MITR LEU fuel design was analyzed using Computational Fluid Dynamics. Coupled conduction and convective heat transfer were modeled for uniform and non-uniform lateral power distributions. It was concluded that, due to conduction, the maximum heat flux ratio on the cladding surface is 1.16, compared to the maximum volumetric power generation ratio of 1.23. The maximum cladding temperature occurs roughly 0.5 inches from the edge of the support plate, while the peak volumetric power generation is located at the end of the fuel meat, about 0.1 inches from the edge of the support plate. Although the heat transfer coefficient is lower in the corner of the coolant channel, this has a negligible effect on the peak cladding temperature, i.e. the peak cladding temperature is related to heat flux only and a "channel average" heat transfer coefficient can be adopted. Moreover, coolant temperatures in the radial direction are reasonably uniform, which is indicative of good lateral mixing. Finally, a quasi-DNS study has been performed to analyze the effect of the fuel grooves on the local heat transfer coefficient. The quasi-DNS results bring useful insights, showing two main effects related to the existence of the grooves. First, the increased surface leads to an increase in the pressure drop and further, the flow aligned configuration of the grooves limits the ability of the near wall turbulent structures to create mixing, leading to a noticeable reduction in the local heat transfer coefficient at the base of the grooves. Overall, this leads to an effective decrease in the local heat transfer coefficient, but due to the increased heat transfer surface the global heat transfer is enhanced in comparison to the flat plate configuration. The improved understanding of the effects of grooves on the local heat transfer phenomena provides a useful contribution to future fuel design considerations. For example, the increase in pressure drop, together with the reduction in the local heat transfer coefficient indicated that the selection of a grooved wall channel instead of a smooth wall channel might not necessarily be optimal, particularly if fabrication issues are taken into account, together with the concern that grooved walls may promote oxide growth and crud formation during the life of the fuel.
Thesis: S.M., Massachusetts Institute of Technology, Department of Nuclear Science and Engineering, 2014.; Cataloged from PDF version of thesis.; Includes bibliographical references (page 85).
2014-01-01T00:00:00ZLoss-of-flow analysis of an unfinned, graded fuel meat, LEU monolithic U-10Mo fuel design in support of the MITR-II fuel conversion
http://hdl.handle.net/1721.1/95600
Loss-of-flow analysis of an unfinned, graded fuel meat, LEU monolithic U-10Mo fuel design in support of the MITR-II fuel conversion
Don, Sarah M
In order to satisfy requirements of the Global Threat Reduction Initiative (GTRI), the 6 MW MIT Research Reactor (MITR-II) is to convert from the current 93%-enr 235U highly-enriched uranium (HEU) fuel to the low-enriched uranium (LEU, <20% 235U) fuel. This reduction in enrichment decreases the neutron flux level due to parasitic absorption by 238U. The neutron flux may be compensated for by increasing the reactor's nominal operating power level to 7.0 MW. Thus a neutronic and thermalhydraulic study was undertaken to evaluate new fuel designs with graded fuel meat thickness and unfinned clad that provide sufficient safety margins for steady-state operation at 7.0 MW. A previously-studied 18-plate LEU fuel design and an identical unfinned fuel design were compared to evaluate the effect of fin removal, demonstrating the need for fuel redesign. A recent feasibility study has shown that a 19-plate, unfinned fuel design with graded fuel meat thicknesses (19B25) provides fuel cycle length and steady-state thermal hydraulic safety margins that meet the design criteria. The objective of this study was to use the RELAP5 MOD3.3 code to confirm the steady-state thermalhydraulic safety margin and to analyze the loss-of-flow (LOF) transient performance of this candidate fuel design. Power distributions obtained for beginning-of-life (BOL), middle-of-life (MOL), and end-of-life (EOL) were analyzed to study the effect of core power distribution and burnup-dependent thermal properties on safety margins. Results show that the MITR-II can safely operate at 7.0 MW with the proposed LEU fuel with an adequate margin (40%) to the onset of nucleate boiling (ONB) -limiting power level. The minimum margin between coolant channel wall and saturation temperatures was at least 22 C in the most limiting channel, in the most limiting core (BOL) at 7.0 MW. The proposed LEU fuel design also performed well during a simulated LOF transient after operation at 7.0 MW, with a peak fuel temperature of 106 C reached in the hot channel, which is well below the U-1OMo blistering temperature of 365*C. During the LOF transient, the maximum clad temperature was 980, meaning that no boiling occurred even during the LOF transient. Bounding analysis to evaluate the effect of an oxide layer and fuel meat thermal conductivity due to fuel burnup estimated that up to a 15 C peak fuel temperature rise can be attributed to increased thermal resistance of oxide layer and fuel thermal conduction reduction. Thus under the most conservative assumption, the estimated peak fuel temperature is 121 C, well under the blistering temperature limit of 365 C. It is concluded that the 19-plate unfinned fuel design with graded fuel meat thickness is a promising candidate for the conversion to LEU fuel and power uprate.
Thesis: S.M., Massachusetts Institute of Technology, Department of Nuclear Science and Engineering, 2014.; Thesis: S.B., Massachusetts Institute of Technology, Department of Nuclear Science and Engineering, 2014.; Cataloged from PDF version of thesis.; Includes bibliographical references (pages 113-115).
2014-01-01T00:00:00ZThe Greedy Exhaustive Dual Binary Swap methodology for fuel loading optimization in PWR reactors using the poropy reactor optimization Tool
http://hdl.handle.net/1721.1/95599
The Greedy Exhaustive Dual Binary Swap methodology for fuel loading optimization in PWR reactors using the poropy reactor optimization Tool
Haugen, Carl C. (Carl Christopher)
This thesis presents the development and analysis of a deterministic optimization scheme termed Greedy Exhaustive Dual Binary Swap for the optimization of nuclear reactor core loading patterns. The goal of this optimization scheme is to emulate the approach taken by an engineer when manually optimizing a reactor core loading pattern. This is to determine if this approach is able to locate high quality patterns that, due to their location in the core loading solution space, are consistently missed by standard stochastic optimization methods such as those in the genetic algorithm class, or those in the simulated annealing class. This optimization study is carried out using the poropy tool to handle the reactor physics model. Initially, optimizations are carried out using beginning of cycle eigenvalue as a surrogate for core excess reactivity and thus cycle length. The deterministic Dual Binary Swap is found to locate acceptable patterns less reliably than stochastic methods, but those that are located are of higher quality. Optimizations of the full depletion problem result in the deterministic Dual Binary Swap optimizer locating patterns that are of higher quality than those found by the stochastic Simulated Annealing, with comparable frequency. The Dual Binary Swap optimizer is, however, found to be very dependent on the starting core configuration, and can not reliably find a high quality pattern from any given starting configuration.
Thesis: S.M., Massachusetts Institute of Technology, Department of Nuclear Science and Engineering, 2014.; Cataloged from PDF version of thesis.; Includes bibliographical references (pages 151-153).
2014-01-01T00:00:00ZImprovements and applications of the Uniform Fission Site method in Monte Carlo
http://hdl.handle.net/1721.1/95598
Improvements and applications of the Uniform Fission Site method in Monte Carlo
Hunter, Jessica Lynn
Monte Carlo methods for reactor analysis have been in development with the eventual goal of full-core analysis. To attain results with reasonable uncertainties, large computational resources are needed. Variance reduction methods have been developed in order to reduce the computational resources required to obtain results in a practical amount of time. This work seeks to expand research in the Uniform Fission Site (UFS) method, a variance reduction technique recently developed that causes uniformity in uncertainty distributions by forcing uniformity in source distributions. This work aims to both improve the method as well as investigate its use with a source acceleration method, Coarse Mesh Finite Difference (CMFD) acceleration. Both techniques have been implemented into OpenMC, a continuous energy Monte Carlo code. The UFS method uses weights to alter the number of neutrons born at a fission site. It operates on a superimposed mesh, in which each mesh cell contains a different weight. These weights use an estimate of the source fraction and fuel volume fraction within the cell to produce uniformity. In current implementations, the fuel volumes are assumed to be dispersed equally over all mesh cells. This work aims to provide an estimate of the fuel volume fraction in each cell in order to improve the accuracy of the method for irregular geometries. The new fuel volume approximation method is tested on a toy problem and on a model of the Advanced Test Reactor, a core with highly irregular geometry. Figures of merit were calculated for a basic Monte Carlo simulation, a simulation with the standard UFS implementation, and the new UFS method with estimated volume fractions. With the toy problem, the new method showed significant improvement and had the highest figure of merit. In the case of the ATR, the long run time for the approximation lowered the figure of merit. Both problems demonstrated that the use of the standard UFS implementation on an irregular geometry produced higher uncertainties than not using the method at all. The UFS method, when used with the estimated volume fractions, behaved as expected and produced uniform uncertainty distributions. The investigation of the use of the UFS method with CMFD acceleration was conducted using the 3-D BEAVRS benchmark. Results showed that keeping CMFD acceleration on during active batches maintained a stationary source and reduced the variance for assembly results. The UFS method stacked on this, reducing the maximum relative uncertainties. The UFS method had variable results with different tallies, but no interference between the two methods was observed.
Thesis: S.M., Massachusetts Institute of Technology, Department of Nuclear Science and Engineering, 2014.; Cataloged from PDF version of thesis.; Includes bibliographical references (pages 61-63).
2014-01-01T00:00:00Z