Theses - Dept. of Nuclear Engineering
http://hdl.handle.net/1721.1/7606
2016-03-23T12:48:17ZOptimization of deep boreholes for disposal of high-level nuclear waste
http://hdl.handle.net/1721.1/97968
Optimization of deep boreholes for disposal of high-level nuclear waste
Bates, Ethan Allen
This work advances the concept of deep borehole disposal (DBD), where spent nuclear fuel (SNF) is isolated at depths of several km in basement rock. Improvements to the engineered components of the DBD concept (e.g., plug, canister, and fill materials) are presented. Reference site parameters and models for radionuclide transport, dose, and cost are developed and coupled to optimize DBD design. A conservative and analytical representation of thermal expansion flow gives vertical velocities of fluids vs. time (and the results are compared against numerical models). When fluid breakthrough occurs rapidly, the chemical transport model is necessary to calculate radionuclide concentrations along the flow path to the surface. The model derived here incorporates conservative assumptions, including instantaneous dissolution of the SNF, high solubility, low sorption, no aquifer or isotopic dilution, and a host rock matrix that is saturated (at a steady state profile) for each radionuclide. For radionuclides that do not decay rapidly, sorb, or reach solubility limitations (e.g., 1-129), molecular diffusion in the host rock (transverse to the flow path) is the primary loss mechanism. The first design basis failure mode (DB 1) assumes the primary flow path is a 1.2 m diameter region with 100x higher permeability than the surrounding rock, while DB2 assumes a 0.1 mm diameter fracture. For the limiting design basis (DB 1), borehole repository design is constrained (via dose limits) by the areal loading of SNF (MTHM/km2 ), which increases linearly with disposal depth. In the final portion of the thesis, total costs (including drilling, site characterization, and emplacement) are minimized ($/kgHM) while borehole depth, disposal zone length, and borehole spacing are varied subject to the performance (maximum dose) constraint. Accounting for a large uncertainty in costs, the optimal design generally lies at the minimum specified disposal depth (assumed to be 1200 in), with disposal zone length of 800-1500 m and borehole spacing of 250-360 meters. Optimized costs range between $45 to $191/kgHM, largely depending on the assumed emplacement method and drilling cost. The best estimate (currently achievable), minimum cost is $134/kgHM, which corresponds to a disposal zone length of -900 meters and borehole spacing of 272 meters.
Thesis: Ph. D., Massachusetts Institute of Technology, Department of Nuclear Science and Engineering, 2015.; Cataloged from PDF version of thesis.; Includes bibliographical references (pages 223-240).
2015-01-01T00:00:00ZInvestigating the educational effectiveness of a science museum exhibit on small modular fusion reactors
http://hdl.handle.net/1721.1/97967
Investigating the educational effectiveness of a science museum exhibit on small modular fusion reactors
Batie, Margo Alexandra
Most people are unaware of the tremendous potential fusion reactors and smaller, more modular reactors possess. To inform them, a science exhibit was.constructed to investigate whether or not it would more effectively teach the audience which in this case are passersby on the first floor of MIT's building 24, about small modular reactors (SMRs) compared to an executive summary written to explain the same technology. Through the employment of hand written surveys, visitor feedback from the executive summary was compared to visitor feedback on the exhibit. The data indicated that although the exhibit lacked the technical detail of the executive summary, it provided a larger proportion of visitors with sufficient background information and a greater appreciation and understanding of fusion energy and reactor modularity. Future SMR exhibits should employ more elements that encourage visitor interaction, such as a demonstration of plasma behavior, as well more information on the cost and feasibility of the technology.
Thesis: S.B., Massachusetts Institute of Technology, Department of Nuclear Science and Engineering, 2014.; Cataloged from PDF version of thesis.; Includes bibliographical references (pages 150-151).
2014-01-01T00:00:00ZInvestigation of sub-meter shields for a low aspect ratio D-T Tokamak fusion reactor
http://hdl.handle.net/1721.1/97966
Investigation of sub-meter shields for a low aspect ratio D-T Tokamak fusion reactor
French, Cameron T
A significant effort is being made by fusion researchers to minimize the total size of magnetic fusion devices on the path toward developing fusion energy. The spherical tokamak, which has a very low aspect ratio, is the most promising of the compact magnetic fusion reactor designs. This compactness imposes a severe material constraint on the design, as a highly compact device will have very thin inner shielding. This inner shielding, which in traditional designs is required to be around 1 meter thick, acts to protect the central solenoid and return toroidal field coil legs from material damage and nuclear heating resulting from high neutron fluxes. The use of a sub-meter inner shield creates potential for the design of a proof of principle magnetic fusion device, sacrificing the central component materials for a demonstration of temporary fusion power production. The nuclear heating of thin shields (~ 0.1 - 0.2m) of various compositions was explored using the Monte Carlo N-Particle (MCNP) transport code. The principal finding was that nuclear heating is the largest concern to the central inboard components. Nuclear heating of these sensitive materials was found to be minimized by the use of a magnesium borohydride blanket with a tungsten first wall. The resulting nuclear heating density for a 100MW, R=1m D-T tokamak employing 0.1 - 0.2m shields is shown to have the potential to threaten the ability of such a device to sustain net electricity.
Thesis: S.B., Massachusetts Institute of Technology, Department of Nuclear Science and Engineering, 2014.; "June 2014." Cataloged from PDF version of thesis.; Includes bibliographical references (page 22).
2014-01-01T00:00:00ZCOMSOL finite-element analysis : residual stress measurement of representative 304L/308L weld in spent fuel storage containers
http://hdl.handle.net/1721.1/97965
COMSOL finite-element analysis : residual stress measurement of representative 304L/308L weld in spent fuel storage containers
Solis, Dominic (Dominic R.)
The ultimate storage destination for spent nuclear fuel in the United States is currently undecided. Spent fuel will be stored indefinitely in dry cask storage systems typically located on-site at the reactor or at a dedicated independent spent fuel storage installation (ISFSI). Since these canisters were not originally designed or qualified for indefinite storage, there is a need to quantify the length of time they will be viable for storing spent fuel. Stress corrosion cracking (SCC) is a concern in these canisters if they are exposed to an aqueous, chloride-containing film. Canisters are fabricated using a concrete overpacking, along with austenitic stainless steel on the inside which is welded together. One factor that would significantly impact SCC behavior inside these canister welds, if the proper conditions developed such that SCC occurred, is the tensile residual stress profile. As the highest residual stresses are present in the welds and their heat-affected zones (HAZ), it would be useful to investigate their influence by predicting the residual stress profile in the container. These data will support further research into the life expectancy of these canisters and the possible ways in which they might fail due to SCC. Residual stress data for nuclear waste canisters are scarce. Without experimental measurements, initial insight must be attained through computational analysis using finite-element analysis (FEA) packages such as COMSOL. Using a representative 304L/308L weld plate as a model in COMSOL, predicted residual stress shows some agreement with expected trends: high tensile stresses in the weld/ HAZ regions and compressive stresses in the surrounding material. Hardness tests show trends similar to the hardening profiles that were created after the weld simulation. Additionally, the thermal model may offer insight in predicting the HAZ profiles in the weld. While the 2D model is simplified and would benefit from further refinement and validation, preliminary results suggest that FEA could be used for residual stress measurement predictions.
Thesis: S.B., Massachusetts Institute of Technology, Department of Nuclear Science and Engineering, 2014.; Cataloged from PDF version of thesis.; Includes bibliographical references (pages 30-31).
2014-01-01T00:00:00Z