Thesis (S.M.)--Massachusetts Institute of Technology, Dept. of Nuclear Science and Engineering, 2012. / "June 2012." Cataloged from PDF version of thesis. / Includes bibliographical references (p. 84-90). / by Saaransh Gulati. / S.M.

Thesis (S.M.)--Massachusetts Institute of Technology, Dept. of Nuclear Science and Engineering, 2010. / "Research funded by the Department of Energy's Advanced Fuel Cycle Initiative Fellowship"--Abstract. Cataloged from PDF version of thesis. / Includes bibliographical references (p. 91-93). / This research, funded by the Department of Energy's Advanced Fuel Cycle Initiative Fellowship, was focused on developing a new approach to studying the nuclear fuel cycle: instead of using the trial and error approach currently used in actinide management studies in which reactors are designed and then their performance is evaluated, the methodology developed here first identified relevant fuel cycle objectives like minimizing decay heat production in a repository, minimizing Pu-239 content in used fuel, etc. and then used optimization to determine the best way to reach these goals. The first half of this research was devoted to identifying optimal flux spectra for irradiating used nuclear fuel from light water reactors to meet fuel cycle objectives like those mentioned above. This was accomplished by applying the simulated annealing optimization methodology to a simple matrix exponential depletion code written in Fortran using cross sections generated from the SCALE code system. Since flux spectra cannot be shaped arbitrarily, the second half of this research applied the same methodology to material composition of fast reactor target assemblies to find optimal designs for minimizing the integrated decay heat production over various timescales. The neutronics calculations were performed using modules from SCALE and ERANOS, a French fast reactor transport code. The results of this project showed that a thermal flux spectrum is much more effective for transmuting used nuclear fuel. In the spectral optimization study, it was found that a thermal flux spectrum is approximately five times more effective at reducing long-term decay heat production than a fast flux spectrum. This conclusion was reinforced by the results of the target assembly material optimization study, which found that by adding an efficient moderator to a target assembly designed for minor actinide transmutation, the amount of decay heat generated over 10,000 years of cooling can be reduced by over 50% through a single pass in a fast reactor without exceeding standard cladding fluence limits. / by Mark Massie. / S.M.

Thesis (Ph. D.)--Massachusetts Institute of Technology, Dept. of Nuclear Science and Engineering, February 2013. / "December 2012." Cataloged from PDF version of thesis. / Includes bibliographical references (p. 145-160). / Processing information quantum mechanically allows the relatively efficient solution of many important problems thought to be intractable on a classical computer. A primary challenge in experimentally implementing a quantum information processor is the control and suppression of environmental noise that decoheres the quantum system and causes it to behave classically. Environmental errors may be dynamically suppressed by applying coherent control pulses to the qubits that decouple the environment. However, the pulses themselves are subject to implementation errors, which hinders the ability to robustly store a complete quantum state. This thesis details results on the use of optimal control theory, noise twirling, and logical qubit encodings to design high-fidelity control pulses and decoupling sequences that are robust to implementation errors. Results are also presented that demonstrate how high-fidelity inductive control of a quantum system may be obtained with limited resonator bandwidth, with a discussion of applications to actuator-based quantum information processors. In a multi-mode design for such a processor, which allows efficient removal of entropy, a new protocol is suggested that permits robust parallel information transfer between nodes. The results detailed in this thesis apply broadly to most implementations of quantum information processing and specifically enable a new design for a spin-based multinode quantum information processor based on single-crystal molecular monolayer electron-nuclear spin systems integrated with superconducting electronics. / by Troy William Borneman. / Ph.D.

Thesis (S.M.)--Massachusetts Institute of Technology, Dept. of Nuclear Science and Engineering, 2012. / Cataloged from PDF version of thesis. / Includes bibliographical references (p. 88-90). / Studies are underway in support of the MIT research reactor (MITR-II) conversion from high enriched Uranium (HEU) to low enriched Uranium (LEU), as required by recent non-proliferation policy. With the same core configuration and similar assembly type, high-density monolithic U-Mo fuel will replace the current HEU fuel with comparable performance. Part of the required analysis for relicensing includes detailed fuel management and burnup studies with the new LEU fuel, to be carried out with a recently developed fuel management tool called MCODE-FM. This code-package is a Python wrapper enabling automatic fuel shuffling between successive runs of MIT's MCODE, which couples MCNP with ORIGEN for full-core neutronics and depletion. In this work, the capabilities of MCODE have been expanded, and the effects of depletion mesh parameters have been explored. Several features have been added to the fuel management tool to encompass the the full range of fuel management options needed for detailed analysis, including assembly flipping, rotation, and temporary storage above the core. In addition, an option to easily manage experiments and custom dummy elements has been added, and a parallel version of MCODE for MCODE-FM that better handles finer discretizations of full-core runs has been developed. These changes have been made in the main wrapper utility as well as the graphical user interface (GUI). In addition to the new MCODE-FM capabilities, a suite of automatic data analysis utilities were developed to consistently parse results. These include utilities to extract or calculate isotope data, fission powers, blade heights, peaking factors, and 3D VTK files for visualization at any time step. The suite has been developed as a series of Python scripts, accessible also through the MCODE-FM GUI. Finally, the effects of the spatial discretization parameters for the depletion mesh have been explored, and mesh choice recommendations have been made for different types of studies. In summary, coarser meshes in the radial and lateral dimensions have been found to yield conservative power peaking results, whereas a finer axial mesh is needed axially. Thus for iterative fuel management studies a fast-running depletion mesh of 8 axial regions, 3 radial regions, and 1 lateral region can be used. However, for safety studies and benchmarking that only need to run once or twice, 16 axial regions, 15 or 18 radial regions (HEU or LEU, respectively), and 4 lateral regions should be used. / by Nicholas E. Horelik. / S.M.

This electronic version was submitted by the student author. The certified thesis is available in the Institute Archives and Special Collections. / Thesis (Ph. D.)--Massachusetts Institute of Technology, Dept. of Nuclear Science and Engineering, 2007. / Includes bibliographical references (p. 245-255). / This study investigates the thermal transport behavior of nanoparticle colloids or nanofluids. The major efforts are: to determine methods to characterize a nanoparticle colloid's mass loading, chemical constituents, particle size, and pH; to determine temperature and loading dependent viscosity and thermal conductivity; to determine convective heat transfer coefficient and viscous pressure losses in an isothermal and heated horizontal tube; and finally to determine the feasibility for potential use as enhanced coolants in energy transport systems, with focus on nuclear application. The efforts result in proving that the two selected nanofluids, alumina in water and zirconia in water, have behavior that can be predicted by existing single phase convective heat transfer coefficient and viscous pressure loss correlations from the literature. The main consideration is that these models must use the measured mixture thermophysical properties. With the acquired knowledge of the experiments, investigation into the potential use or optimization of a nanofluid as an enhanced coolant is further explored. The ultimate goal of contributing to the understanding of the mechanisms of nanoparticle colloid behavior, as well as, to broaden the experimental database of these new heat transfer media is fulfilled. / by Wesley Charles Williams. / Ph.D.

Thesis (S.B.)--Massachusetts Institute of Technology, Dept. of Nuclear Science and Engineering, 2012. / "June 2012." Cataloged from PDF version of thesis. / Includes bibliographical references (p. 50-55). / The danger of high, acute doses of radiation is well documented, but the effects of low-dose radiation below 100 mSv is still heavily debated. Four theories concerning the effects of lowdose radiation are presented here: supra-linearity, linear-no-threshold (LNT), threshold, and hormesis. The available evidence for and against these theories, which falls into the categories of either epidemiological studies, in vitro cell experiments, or in vivo animal experiments, includes studies which support each of the four theories. Currently, all radiation risk estimates are based on an LNT interpretation of the life span study (LSS) of atomic bomb survivors in Japan. However, while this pattern is undisputed at high doses, this linear extrapolation of risk to low doses is challenged by many recent experiments involving cell mechanisms and animal models, and there is also high uncertainty involved in estimating risk using only epidemiological studies.. Variations have also been observed depending on dose-rate, the organ at risk, and other factors for which the current data cannot adequately account. While the evidence is still inconclusive, the existence of a threshold in human responses to low-dose radiation would drastically alter current guidelines, such as those currently restricting many people from returning to their hometowna in Fukushima, Japan. Thus, it is important to further investigate these low*dose responses in order to more fully describe the risks and to create more accurate radiation guidelines. / by Elizabeth J. Wei. / S.B.

Thesis (S.B.)--Massachusetts Institute of Technology, Dept. of Nuclear Science and Engineering, 2012. / "June 2012." Cataloged from PDF version of thesis. / Includes bibliographical references (p. 29). / A study of the prompt losses of injected neutral beam born fast ions was conducted on the DIII-D tokamak at General Atomics using scintillator based fast ion loss detectors (FILD) and a reverse orbit calculation code. Prompt losses, also called first orbit losses, result from injected neutrals that are ionized on orbits that terminate to the outer wall before making a complete neoclassical, poloidal revolution. A strike map code has been developed which generates meshes that overlay optical fast ion signals from the FILD scintillator, providing a measurement of the pitch angles and gyroradii of incident fast ions. The pitch angles and gyroradii of incident ions are inputs to a reverse orbit calculation code used to calculate the trajectories of the incident ions in reverse time back to their birth at the intersection of the reverse orbit and an overlaid neutral beam injection footprint. The megahertz (MHz) sampling frequency of the FILD scintillator, along with finer time resolution neutral beam signals, enabled a comparison of the measured time delay between the onset of the neutral beam injection and the measured FILD loss signals with the calculated transit time based on the path length of the simulated reverse orbit. Consistency between the experimentally measured transit times and the simulation orbit times was observed. This result indicates the generated strike maps which provide a measurement of incident ions' gyroradii and pitch angles are accurate. This study supplements current studies seeking to improve the understanding of fast ion transport due to magnetohydrodynamic (MHD) activity, such as reverse shear Alfven eigenmodes (RSAEs) and toroidal Alfven eigenmodes (TAEs), which will be of great importance for predominately self-heated reactor scenarios. / by Derek A. Sutherland. / S.B.

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). / 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. / by Ethan Allen Bates. / Ph. D.

Thesis (Ph. D.)--Massachusetts Institute of Technology, Dept. of Nuclear Science and Engineering, 2011. / Cataloged from PDF version of thesis. "June 2011." / Includes bibliographical references (p. 180-184). / This thesis presents the implications of using a risk-informed licensing framework to inform the design of Sodium Fast Reactors. NUREG-1860, more commonly known as the Technology Neutral Framework (TNF), is a risk-informed licensing process drafted by the Nuclear Regulatory Commission's (NRC) Office of Nuclear Regulatory Research. The TNF determines the acceptability of accident sequences by examining the 95th percentile estimate of both the frequency and quantity of radioactive material release and compares this value to predetermined limits on a Frequency-Consequence Curve. In order to apply this framework, two generic pool type sodium reactors, one using metal fuel and one using oxide fuel, were modeled in RELAP5-3D in order to determine the transient response of reactors to unprotected transient overpower and unprotected loss of flow events. Important transient characteristics, such as the reactivity coefficients, were treated as random variables which determine the success or failure of surviving the transient. In this context, success is defined as the cladding remaining intact and the avoidance of sodium boiling. In order to avoid running an excessive amount of simulations, the epistemic uncertainties around the random variables are sampled using importance sampling. For metallic fuel, the rate of fuel/cladding eutectic formation has typically been modeled as an Arrhenius process which depends only on the temperature of the fuel/cladding interface. Between the 1960s and the 1990s, numerous experiments have been conducted which indicate that the rate of fuel/cladding eutectic formation is more complex, depending upon fuel/cladding interfacial temperature, fuel constituents (uranium metal or uranium zirconium), cladding type (stainless steel 316, D9 or HT9), linear power, plutonium enrichment and burnup. This thesis improves the modeling accuracy of eutectic formation through the application of multivariable regression using a database of fuel/cladding eutectic experiments and determines that the remaining uncertainty governing the rate of eutectic formation should not significantly affect the frequency of cladding failure for tested cladding options. The general conclusion from this thesis is that when using NUREG-1860 to license metal or oxide fueled SFRs, it is steady state, not transient, cladding considerations which control optimal operating temperature, currently corresponding to an approximate core outlet temperature of 550°C. Metallic cores traditionally have been designed with core outlet temperatures of 510°C and increasing this temperature to 550°C may decrease the busbar cost by 19% when combined with the adoption of a Supercritical-CO₂ power conversion cycle, reduced containment requirements, and Printed Circuit Heat Exchangers. While both fuel types will be shown to meet the NUREG-1860 requirements, the frequency of radiation release for unprotected loss of flow and unprotected transient overpower events for metallic fuel has been shown to be orders of magnitude lower than for oxide fuel. / by Matthew R. Denman. / Ph.D.

Thesis: S.M., Massachusetts Institute of Technology, Department of Nuclear Science and Engineering, 2018. / This electronic version was submitted by the student author. The certified thesis is available in the Institute Archives and Special Collections. / Cataloged from student-submitted PDF version of thesis. / Includes bibliographical references (pages 89-104). / The economic potential and readiness of various cross-cutting technologies, both nuclear-related and not, have been evaluated to find that a 28-38% net reduction in LCOE is possible for new nuclear builds beginning construction before 2030, and up to 65% after that. The short term benefit comes primarily from several capital cost reducing technologies such as seismic isolation (7-10%), steel plate composites and ultra-high performance concrete (6-8%), modular construction of mechanical components (4-5%), and high-strength rebar (~2%), with heat transfer coatings (~5%) being the only non-capital technology to have a comparable impact. The long term benefit is also predominantly due to capital technologies, with the additive manufacturing of large metal components (3-9%) and offshore siting (3-9%) accounting for most of the increased benefit. Existing sites likewise look to profit, with retrofits enabling savings equivalent to a 6-8% reduction in LCOE in the near term - principally because of the aforementioned coatings. Other technologies evaluated include accident tolerant fuels, advanced instrumentation and control, advanced power cycles, embedment, energy storage, and robotics. As plant overnight costs and construction times have tripled in the US since Three Mile Island, such technologies thus have significant potential to aid a troubled industry. These estimates notably do not include learning from accumulated construction experience, which can additionally account for up to a 20-40% reduction in LCOE and is a driving incentive of small modular reactors, or revenue enhancement from sources such as the secondary objectives of Brayton cycles, which were found to be the most likely motivation in selecting such alternatives, and energy storage, where thermal storage was identified to be the best fit for nuclear plants. Furthermore, the durability of heat transfer coatings was determined to be more important to their viability than thermal performance, once past a relatively low threshold. And though the above values define the feasible range of savings, care must be taken in implementation to achieve these savings. Issues with modular construction, for example, can arise if too much of it is implemented too quickly as it is generally less flexible than traditional construction. This problem was observed at recent US AP1000 builds. / by Patrick A. Champlin. / S.M.