Thesis (Ph. D.)--Massachusetts Institute of Technology, Dept. of Nuclear Science and Engineering, 2010. / Cataloged from PDF version of thesis. / Includes bibliographical references (p. 285-295). / In order to significantly increase the power density of Light Water Reactors (LWRs), the helical-cruciform (HC) fuel rod assembly has been proposed as an alternative to traditional fuel geometry. The HC assembly is a self-supporting nuclear fuel configuration consisting of 4-finned, axially-twisted fuel rods closely packed against one another in a square array. Within the LWR core, HC fuel would in theory possess several inherent advantages over traditional fuel, potentially allowing for operation at a higher power density. Chief among these advantages are a larger surface-to-volume ratio, a shorter radial heat conduction path, and improved mixing characteristics. In previous work, computational models of the HC fuel assembly have been of limited accuracy due to the absence suitable correlations. To address needs within these subchannel analysis models, experimental measurements of rod bundle coolant mixing have been conducted with 4x4 arrays of HC test rods. The tests used the technique of a hot water tracer injection (at 95°C) into a bulk flow of cold water (at 25°C). Downstream temperature measurements were used to judge the rate of lateral cross-flow within the HC rod bundle. These tests were conducted at atmospheric pressure, and encompassed a range of mass fluxes from 1000 kg/m2s to 3500 kg/m2s, HC rod twist pitches of 200cm, 100cm, and 50cm, and different hot water injection velocities and mixing lengths. Data from over 300 tests was analyzed, yielding a best fit correlation for use with any twist pitch, rod length, or coolant flow rate. Compared to the bare rod bundle, this correlation implies an enhancement in the intensity of turbulent interchange of 40% brought about by the HC geometry, and a 1.6% forced diversion of axial flow per subchannel, per quarter-turn along the rod length. These parameters fit all data points considered within a standard deviation of 24%. Stochastic error was limited to ±16% by the use of precise temperature sensors. By applying this empirical mixing model to the subchannel representation of a BWR core featuring the HC rod design, a need to increase the flow area of the edge subchannels was demonstrated. This prompted a slight re-design of the HC fuel rod cross-section in order to make room for small spacer protrusions at the duct wall, to increase flow to peripheral subchannels. The modification was accomplished by reducing fin length, but increasing the inner diameter to maintain the reference fuel volume. The water rod region was also adjusted to maintain the reference assembly hydrogen to uranium atom ratio. With this modification, the model predicted a 24% allowable power uprate for the 200cm twist pitch HC core. Inlet and exit enthalpies were maintained from the reference cylindrical-rod core. When applied to a PWR core of HC rods, also with a fixed power to flow ratio, this empirical mixing model predicted an allowable power uprate of 47%, using traditional CHF correlations for cylindrical fuel. In subcooled conditions, CHF is known to be more sensitive to peaked areas of non-uniform heat-flux than in saturated two-phase flow conditions. Therefore power density gains will likely be dependent on the degree to which the rod twist would disrupt of nascent pockets of vapor; this effect should be further investigated experimentally. In order to further ascertain the potential gain in power density for the new design, an experiment must be carried out to obtain CHF data for the HC rod bundle. Two facilities with this aim were designed in great detail for BWR conditions: the first would operate using high pressure water at 7MPa, and the alternate would use a relatively low pressure refrigerant at equivalent conditions. The appropriate scaling laws were applied, which resulted in the choice of R134a as the simulant fluid. The R134a facility was found to be possible to construct at a greatly reduced cost. / by Thomas M. Conboy. / 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. 46-47). / The key difference between a regulated and a liberalized electricity market is the establishment of a competitive generation marketplace via spot markets, day-ahead auctions, and over-the-counter trading activity. In a liberalized market, power plants are no longer guaranteed a fixed return on capital investments or the ability to pass on increases in fuel prices to customers directly. Therefore, power generators have had to modify their capital allocation and marketing strategies to resemble that of a typical competitive market participant more closely, balancing expected returns with portfolio risk. Advanced Combined Cycle Gas Turbine (CCGT) power plants are currently viewed as the most attractive generation investment option, offering low capital costs, short construction lead-times and financial optionlike qualities. In contrast, a nuclear power plant's levelized cost is dominated by large fixed costs and capital expenditures. Even the perception of nuclear power as being a hedge against volatile natural gas markets has been called into question by power market Monte Carlo simulations. These simulations indicate that CCGT power plants are actually the generation option with the least exposure to natural gas and electricity price uncertainty because of the intrinsic hedge created by the historically high correlation of natural gas and electricity prices[1, 2]. Nevertheless our simulations, heavily focused on modeling the non-linearity of the power supply curve, indicate that the portfolio diversification value of nuclear power is dependent on the generation composition of the power market. In markets primarily composed of natural gas fired capacity, an investment in nuclear power offers no portfolio diversification value, with all three baseload generation types are effectively long positions in natural gas. Conversely, in markets with a large amount of coal capacity there is a competition for market share between major marginal fuel types, coal and natural gas, which creates less favorable market dynamics for the CCGT. While we still observe a high natural gas-electricity correlation, the intrinsic hedge no longer stabilizes the CCGT profits. Our simulations indicate that in a bi-marginal fuel market a CCGT power plant is short natural gas, with cheaper natural gas helping to boost capacity factors, reduce operational heat rates, and displace coal power plants. Similarly, as currently observed in Northeastern power markets, cheap natural gas has not only shrunk coal power profit margins but also negatively impacted plant capacity factors. Therefore, the portfolio diversification value of nuclear comes from it being insulated from fossil fuel price uncertainty, but not because this attribute equates to a more stable levelized cost. Rather, nuclear power's low cost and low volatility fuel insures that an unfavorable shift in fossil fuel prices will not result in a large decrease in capacity factor and subsequent increase in profit volatility. / by Malcolm Bean. / S.M.

Thesis (S.B.)--Massachusetts Institute of Technology, Dept. of Nuclear Science and Engineering, 2006. / Includes bibliographical references (p. 51-52). / The four biggest energy sources in the United States are coal, crude oil, natural gas, and nuclear power. While coal and nuclear power are produced domestically, more than 70% of crude oil and 20% of natural gas is imported. This places an unhealthy dependence on foreign products for our economy. Just as importantly, all of these energy sources, with the exception of nuclear power, produce large amounts of polluting emissions in the form of greenhouse gases which are responsible for environmental degradation. For these two reasons, we explore possible government policies to shift the US energy economy towards domestically-produced, environmentally-clean alternative energy sources, the most prevalent of which is nuclear power. Different forms of government support for investment in nuclear power is discussed, such as investment tax credits and production tax credits. As an instrument of public policy to affect energy imports and environmental impact, the possibility of a carbon tax (on the order of $150/tC) is considered. The effects of this carbon tax on the energy sector in the medium-term future (in the year 2020) are analyzed. Under the constraint of maintaining current natural gas demand the results show that there will be an increase in the use of nuclear power while lowering the dependence on crude oil and coal. To accomplish this, the use of natural gas is shifted from the power sector to the residential, commercial and industrial sectors due to the economic incentives to do so. From an environmental perspective, this carbon tax lowers emissions by a predicted 30% of its 2020 business-as-usual rates. Economically, the carbon tax lowers crude oil import levels by 20% and reduces the US balance of payments by over $170 billion in the year 2020. / by Tanzeer S. Khan. / S.B.

Thesis (S.B.)--Massachusetts Institute of Technology, Dept. of Nuclear Science and Engineering, June 2010. / "June 2010." Cataloged from PDF version of thesis. / Includes bibliographical references (p. 28). / With growing demands for commercial nuclear power, there is also a growing need for better energy efficiency from nuclear power reactors. In order to reach a high burnup up to 100 MWd/kg, previous research has examined the use of thorium-plutonium mixedoxide fuel as a potential candidate for this high-burnup goal. Though the neutronics studies have looked upon this fuel type favorably, the purpose of this paper is to investigate the self-protection capabilities of this fuel type, for anti-proliferation purposes. In particular, there were two proliferation-resistance methods that were analyzed. First, this study examined the time-dependant dose-rate of the spent fuel caused by the decay of the isotope uranium-232, which releases a high-energy gamma of 2.6 MeV. Next, this study examined the possibility of denaturing the fuel with depleted uranium in order to dilute the weapons-usable isotope uranium-233 in the spent fuel. The U-232 dose rate was also calculated for the denatured case. Ultimately, the study found that there was a negligible different in the amount of time that it takes for either fuel type to become self-protective. The denatured case showed that it requires much more plutonium than the undenatured case in order to ensure that there is not sufficient weapons-usable U-233 in the discharged fuel. / by Luis Alberto Torres. / S.B.

Thesis (S.M.)--Massachusetts Institute of Technology, Dept. of Nuclear Science and Engineering, 2007. / Includes bibliographical references (p. 135). / The internally and externally cooled annular fuel is a new type of fuel for PWRs that enables an increase in core power density by 50% within the same or better safety margins as the traditional solid fuel. Each annular fuel assembly of the same side dimensions as the solid fuel has 160 annular fuel rods arranged in a 13x13 array. Even at the much higher power density, the fuel exhibits substantially lower temperatures and a MDNBR margin comparable to that of the traditional solid fuel at nominal (100%) power. The major motivation for such an up-rate is reduction of electricity generation cost. Indeed, the capital cost per kWh(e) of the construction is smaller than the standard construction of a new reactor with solid fuel. Elaborating on previous work, we study the economic payoff of such an up-rate of an existing PWR given the expected cost of equipment and also cost of money using different assumptions. Especially, the fate of the already bought solid fuel is investigated. It is demonstrated that the highest return on investment is obtained by gradually loading annular fuel in the reactor core such that right before shutting the reactor down for the up-rate construction, two batches in the core are of annular fuel. This option implies running a core with a mixture of both annular fuel and solid fuel assemblies. In order to prove the technical feasibility of such an option, the thermal-hydraulics of this mixed core is investigated and the Minimum Departure From Nucleate Boiling is found to be either unaffected or even improved by using a mixed core. Consequently, a neutronic model is developped to verify and validate the neutronic feasibility of the transition from solid fuel to annular fuel. / (cont.) The overall conclusion of this work is that annular fuel is a very promising option for existing reactors to increase by 50% their power, because it enables such an uprate at very attractive return on investement. We show that, by a smart management of the transition, a return on investment of about 22 to 27 % can be achieved. / by Julien Beccherle. / S.M.

Thesis (S.B.)--Massachusetts Institute of Technology, Dept. of Nuclear Science and Engineering, 2007. / Includes bibliographical references. / The lead-cooled flexible conversion ratio fast reactor shows many benefits over other fast-reactor designs; however, the higher power rating and denser primary coolant present difficulties for the design of a passive decay heat removal system. In order to achieve passive cooling, enhancements are needed over current designs, such as the S-PRISM and ABR, which utilize passive cooling through the reactor vessel to atmospheric air. Enhancements such as axial fins, a perforated plate, and round indentations, or dimples, were considered as additions to the hot air riser to increase heat transfer. Other enhancements include a liquid metal bond between the reactor and guard vessels, and a dual-level design which introduces ambient temperature air halfway up the vessel wall. A code was written in Java to simulate these conditions, leading to a promising case using dimples on the guard vessel wall as the primary mode of heat transfer enhancement, and including the dual-level design. A conservative estimate of dimple performance indicates that during a passive decay heat removal shutdown, bulk primary coolant temperature will peak at 713 'C, giving a 12 OC margin to clad failure. Attempts were made to refine the uncertainty within the calculations using a computational fluid dynamics code, Fluent, but these ultimately were unsuccessful. Additional studies were conducted on the static stress imparted on the vessel, and the dynamic stress caused by a seismic event. The static stress was found to be within ASME code limits. Seismic analysis determined that a seismic isolation scheme would be necessary in order to prevent damage to the vessel during an earthquake. / by Joshua Whitman. / S.B.

Thesis: S.B., Massachusetts Institute of Technology, Department of Nuclear Science and Engineering, February 2015. / Cataloged from PDF version of thesis. "February 2015." / Includes bibliographical references (pages 46-47). / Fusion has been "forty years away", that is, forty years to implementation, ever since the idea of harnessing energy from a fusion reactor was conceived in the 1950s. In reality, however, it has yet to become a viable energy source. Fusion's promise and failure are both investigated by reviewing the history of the United States domestic fusion program and comparing technological forecasting by fusion scientists, fusion program budget plans, and fusion program budget history. It is evident that delays in progress were due to both technologic and economic setbacks. In order for the US to become a leader in fusion energy, it must continue supporting domestic fusion experiments while maintaining involvement in ITER. / by Lauren A. Merriman. / S.B.

Thesis (S.B.)--Massachusetts Institute of Technology, Dept. of Nuclear Science and Engineering, 2007. / Includes bibliographical references (p. 57-60). / The subject of this thesis is devoted to a class of multiparticle entangled states known as the cluster-states. In particular, we focused on a system of four spins and studied the entanglement properties of a four-qubit cluster-state, using a set of entanglement measures for quantifying multipartite entanglement. We then experimentally prepared the linear cluster-state in a liquid NMR sample of crotonic acid, by applying a set of pulses generated by the Gradient Ascent Pulse Engineering (GRAPE) algorithm on a temporally averaged pseudo-pure state of four carbon spins. While our spectral results were consistent with the creation of a linear cluster-state, the reconstruction of the experimental density matrix via a full state tomography of the system revealed additional challenges in the detection of certain desired spin terms. These problems must be overcome before the system could be studied quantitatively. / by Jennifer T. Choy. / S.B.

Thesis: Ph. D., Massachusetts Institute of Technology, Department of Nuclear Science and Engineering, 2017. / 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 314-321). / The ability to model and simulate nuclear reactors during steady state and transient conditions is important for designing efficient and safe nuclear power systems. The accurate simulation of a nuclear reactor is particularly challenging because the multiple physical processes within the reactor are tightly coupled, which requires that the numerical methods used to resolve each physical process can accurately and efficiently transfer and utilize data from other applications. Monte Carlo methods are desirable for solving the neutron transport equation required in reactor analysis because of the inherent accuracy of the method, but the Computational Solid Geometry (CSG) representation of the physical geometry makes it difficult to accurately and efficiently perform multiphysics reactor analyses with other applications that utilize finite element or finite volume representations. To address this limitation, a multiphysics coupling framework that minimizes the need for spatial discretization in the Monte Carlo geometry is presented in this thesis. The coupling framework uses Functional Expansion Tallies to transfer multiphysics information from the Monte Carlo application to other multiphysics tools. Additionally, the coupling framework uses a modified method for transporting neutrons through spatially continuous total macroscopic cross section distributions in order to incorporate continuous multiphysics feedback fields such as fuel temperature and coolant density into the Monte Carlo simulation. It has been shown that separable Zernike and Legendre Function Expansion Tallies can effectively reconstruct a continuous distribution of fission power density. Additionally, using a prototypical three-dimensional Light Water Reactor pin cell, the method used to transport neutrons through a continuously varying fuel temperature and coolant density distribution was shown to be 1.7 times faster than a comparable discretized simulation with volume-averaged properties, while still providing a high level of accuracy. Finally, in order to make the overall multiphysics coupling scheme useful for reactor analyses, a novel spatially continuous depletion methodology was developed and investigated. With the spatially continuous depletion methodology, number densities can be represented as a linear combination of polynomials, and those polynomial representations can be integrated through time to predict reactor operation. The spatially continuous depletion methodology was able to accurately predict the eigenvalue and number density distributions in a two-dimensional LWR pin cell depletion containing Gd-157 from a 2 weight percent GdO2 and seven other nuclides in the depletion matrix. Analyses of the spatially continuous depletion methodology showed that significant reductions in the number of tallied values could be achieved if polynomial representations were optimized for each nuclide reaction rate. From the depletion simulations in this thesis, a 23% reduction in the required number of reaction rate tallies compared to a lower-fidelity, 10 radial ring pin discretization was shown to be achievable with nuclide polynomial optimization. In addition to showing potential for reductions in tally memory and computational requirements, the spatially continuous depletion simulation was shown to be equal in computational performance to a discrete simulation with 10 radial rings and 8 azimuthal cuts, while providing a much higher level of spatial fidelity in number density concentrations. / by Matthew Shawn Ellis. / Ph. D.

Thesis: Ph. D., Massachusetts Institute of Technology, Department of Nuclear Science and Engineering, 2017. / 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 448-467). / The nuclear industry is no exception when it comes to those affected by fouling deposit problems. Fouling deposits on fuel rods in nuclear reactors, known as crud, can cause a variety of undesirable effects including axial power shifts, accelerated corrosion, increased primary circuit radiation dose, and possible fuel failure. This study revisits the crud problem once again using a newly constructed Internally Heated Testloop For PWRs (IHTFP) and new analytical techniques, and attempt to find a way to prevent or mitigate crud, or at least better understand it. This is the first time that fuelrod coatings are examined as a way of countering crud growth. These coatings are chosen based on their surface chemical properties and robustness at PWR conditions. For the goal of gaining a better understanding of crud, this study is the first to apply fractal analysis to characterize crud. To achieve both of these goals, the IHTFP was built to obtain crud grown under the PWR thermal-hydraulic and chemical conditions. The crud-resistant coatings experiments show significantly reduced crud surface coverage, indicating reduced crud adhesion, for TiC and ZrN coatings. The results roughly agree with London-van der Waals theoretical force predictions, suggesting that London-VDW forces are responsible for the adhesion of crud to fuel cladding. This knowledge can be useful in designing better crud-resistant materials. The fractal analysis can provide a simple, effective way to characterize the macro-scale behavior of crud with its micro-scale properties. The fractal analysis experimental study found R2 values to be very close to one when applying the box-counting method to crud, which is one piece of evidence to support the usage of fractal analysis on crud. Moreover, a strong logarithmic relationship trend between fractal dimension and porosity was found. This relationship applies to both the IHTFP's and Westinghouse loop's crud, even though the two experimental setups used different crud precursors and heat flux. This could indicate that crud's fractal dimension is dependent only on porosity. This relationship could simplify crud modeling and lead to better predictions of crud's behaviors. Better predictions can lower margins, leading to more efficient reactors. / by Ittinop Dumnernchanvanit. / Ph. D.