Thesis (S.B.)--Massachusetts Institute of Technology, Dept. of Nuclear Science and Engineering, 2007. / "June 2007." Leaves unnumbered. / Includes bibliographical references (leaves [58]-[59]). / In order to explore the physical basis by which high resolution diffusion imaging derives information about fiber alignment and diameter, we simulated a model diffusion experiment employing a random walk paradigm. A simulation of a model diffusion spectrum imaging experiment was written in Java in order to compare the diffusive behavior of particles in a perfectly reflecting channel with the internal compartment of the myofibers contained in an imaged mouse tongue. The simulated probability distribution function (PDF) for diffusion was specifically employed to estimate the myofiber diameter for the cells imaged by tissue imaging of the tongue by DSI. Our group performed a DSI experiment on a mouse tongue with a 4.7 Tesla MRI spin echo experiment in order to reconstruct a set of PDFs for the diameter, which closely correlated with the actual values for cell diameter obtained by 3D microscopic visualization through two photon microscopy of the same tissue. These results provide method for estimating myofiber diameter through the properties of the diffusion PDF obtained by whole tissue magnetic resonance imaging. / by Victor Christopher Cabral. / S.B.

Thesis (S.B.)--Massachusetts Institute of Technology, Dept. of Nuclear Science and Engineering, 2012. / Cataloged from PDF version of thesis. / Includes bibliographical references (p. 38-40). / In order to promote nuclear power production as an attractive option for power generation, measures must be taken to ensure that the process is both safe and economical. One aspect of the nuclear fuel cycle that contributes significantly towards such goals is the management of the fuel. Proper fuel management within a reactor core requires an understanding of the trade-offs between maximizing the reactivity, while simultaneously minimizing the power peaking. To accomplish this, various algorithmic methods can be used to help determine the optimal arrangement of fuel bundles in the core. By exploiting innovative computational tools for the analyses of reactor cores, more accurate and precise calculations can be made, and nuclear power generation can continue to be safe and economical. While traditionally, heuristic algorithms have been used for these large optimization problems, more direct algorithms may have the potential to provide the most favorable configuration for the reactor core. The use of Greedy Exhaustive Dual Binary Swaps (GEDBS) was implemented for the optimization of the quarter-core of a reactor containing 193 fuel assemblies. The primary objective was to investigate the quality of the output from the GEDBS algorithm and to compare it to those of other algorithmic methods. In this way, conclusions could be made as to whether or not the end results from this exhaustive algorithm justified the added runtime associated with their use, and ultimately, help to determine its viability for future research. / by Jessica L. Hammond. / S.B.

Thesis: Ph. D., Massachusetts Institute of Technology, Department of Nuclear Science and Engineering, 2014. / Cataloged from PDF version of thesis. / Includes bibliographical references (pages 209-216). / In an effort to illuminate the effects of the strong plasma gradients in the pedestal region on impurity transport, research was conducted to measure complete sets of impurity density, poloidal and parallel velocity, and temperature at two separate poloidal locations in the pedestal region of the Alcator C-Mod tokamak. To this end, the diagnostic technique gas puff-CXRS was refined and expanded on, allowing for the first time in a tokamak complete measurements of impurities at the high-field side (HFS). Large in-out B5+ impurity density asymmetries were measured in H-mode plasmas with strong boundary electron density gradients, with a build-up of impurity density at the HFS. Impurity temperatures were also found to be asymmetric in the pedestal region, with larger temperatures at the low-field side (LFS). Such temperature asymmetries suggest a significant asymmetry in electron density near the separatrix. In contrast to these H-mode results, plasmas with low boundary electron density gradients, such as L-mode and I-mode, exhibit constant impurity density on a flux surface, even if strong electron temperature gradients are present. Mechanisms which could drive such poloidal asymmetries are explored. Experiments provide evidence against localized impurity sources and fluctuation-induced transport as primary causes. Particle transport timescales are compared, showing that the radial transport becomes comparable to or faster than the parallel transport in the pedestal region. Additionally, modelling of impurity transport using conventional, one-dimensional neoclassical physics fails to correctly reproduce the measured flux-surface averaged impurity density, suggesting along with the timescale estimates that a more complete two-dimensional treatment of impurity particle transport is required. The measured impurity velocities at the LFS and HFS are compared to the canonical form for particle flow velocity within the flux surface of a tokamak. Within the error bars of the measurement, agreement is found with the canonical form. The implications of exact matches to the canonical form are low radial transport, and the E x B drift dominating the perpendicular impurity flow. Further work is motivated into more precise velocity measurements to determine if the velocities exactly match this canonical form. / by Randy Michael Churchill. / Ph. D.

Thesis (S.M. and Nucl. E.)--Massachusetts Institute of Technology, Dept. of Nuclear Science and Engineering, 2007. / "September 2007." / Includes bibliographical references (p. 91-94). / Recently, attention has been drawn to the viability of using S-CO₂ as a working fluid in modem reactor designs. Near the critical point, CO₂ has a rapid rise in density allowing a significant reduction in the compressor work of a closed Brayton Cycle. Therefore > 45% efficiency can be achieved at much more moderate temperatures than is optimal for the helium Brayton cycles. An additional benefit of the S-CO₂ system is its universal applicability as an indirect secondary Power Conversion System (PCS) coupled to most GEN-IV concept reactors, as well as fusion reactors. The United States DOE's GNEP is now focusing on the liquid Na cooled primary as an alternative to conventional Rankine steam cycles. This primary would also benefit from being coupled to an S-CO₂ PCS. Despite current progress on designing the S-CO₂ PCS, little work has focused on the principal supporting systems required. Many of the required auxiliary systems are similar to those used in other nuclear or fossil-fired units; others have specialized requirements when CO₂ is used as the working fluid, and are therefore given attention in this thesis. Auxiliary systems analyzed within this thesis are restricted to those specific to using CO₂ as the working fluid. Particular systems discussed include Coolant Make-up and Storage, Coolant Purification, and Coolant Leak Detection. Concepts discussed include: potential forms of coolant storage, including cryogenic and high pressure gas, with some "back of the envelope" methods which can be used for estimating the coolant transferred; possible coolant contaminants and their sources; options for the procurement of the CO₂ from potential distributors, including available purities and estimated cost; the purity of CO₂ for the S-CO₂ system and purification methods; various methods of coolant leak detection using both insitu analyzers and portable devices for maintenance personnel, and instrumentation for the monitoring of compartmental CO₂ and CO concentrations to meet OSHA standards. / (cont.) A conceptual design is presented for coolant storage. Systems are discussed in terms of basic functionality, system requirements, desired features, basic safety and design concerns, and identification of issues to be resolved by future research. / by Rosemary M. Freas. / S.M.and Nucl.E.

Thesis (Sc. D. in Applied Plasma Physics)--Massachusetts Institute of Technology, Dept. of Nuclear Science and Engineering, 2011. / 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 (p. 183-191). / Electromagnetic plasma waves in the ion cyclotron range of frequencies (ICRF) are routinely used in magnetic fusion experiments to heat plasmas and drive currents. However, many experiments have revealed that wave energy losses in the plasma edge and at the wall are significant, and detected that the acceleration of ions into the walls due to the formation of radio-frequency (RF) sheaths is one of the root causes of this problem. Since the RF-enhanced sheaths have many undesirable effects, such as impurity production and hot spot generation, a predictive numerical tool is required to quantitatively evaluate these effects with complicated boundary shapes of tokamaks taken into account. In this thesis the numerical code that solves self-consistent RF sheath-plasma interactions in the scrape-off layer for ICRF heating is developed based on a nonlinear finite element technique and is applied to various problems in the one-dimensional (1D) and two-dimensional (2D) domains corresponding to simplified models for the poloidal plane of a tokamak. The present code solves for plasma waves based on the cold plasma model subject to the sheath boundary condition, in which the most important physics that happens in the sheath is captured without using the field quantities in the sheath. Using the developed finite element code, several new properties of the RF sheath plasma interactions are discovered. First, it is found in the 1D domain that multiple roots can be present due to the resonance of the propagating slow wave and its nonlinear interaction with the sheath. Second, sheath-plasma waves are identified in a 2D slab geometry, and it is proved in conjunction with an electrostatic 2D sheath mode analysis that the sheath-plasma wave only appears in the vicinity of the sheath surface if the plasma density is greater than the lower hybrid density, and its wavelength depends on various parameters. Third, as a consequence of the self-consistent interaction between the propagating slow wave and the sheath, it is shown that the electric field distribution pattern in the plasma smoothly varies along the magnetic field lines between the conducting-wall and quasi-insulating limits. In the numerical analysis employing the 2D domain whose scale is equivalent to the Alcator C-Mod device, it is demonstrated that the calculated sheath potential can reach the order of kV, which is sufficient to yield enhanced sputtering at the wall. In addition, it is shown that the sheath potential in the close vicinity of the antenna current strap can be insensitive to the direction of the background magnetic field in the RF sheath dominated regime. Further, it is found from a series of nonlinear calculations that the sheath potential sensitively varies depending on the plasma density and electron temperature, which is consistent with the scaling derived from the Child-Langmuir law and the definition of the RF sheath potential. Lastly, a new finite element approach, which is named the finite element wave-packet method, is developed for the purpose of solving for multiscale plasma waves in the tokamak poloidal plane accurately with reasonable computational cost. This method is established by combining the advantages of the finite element and spectral methods, so that important properties in the finite element method, such as the sparsity of the global matrix and the ease in satisfying the boundary conditions, are retained. The present scheme is applied to some illustrative 1D multiscale problems, and its accuracy improvement is demonstrated through comparisons with the conventional finite element method. / by Haruhiko Kohno. / Sc.D.in Applied Plasma Physics

Thesis (S.M.)--Massachusetts Institute of Technology, Dept. of Nuclear Science and Engineering, 2009. / Cataloged from PDF version of thesis. / Includes bibliographical references (p. 73-75). / A 1 OOOMWth liquid-salt cooled thermal spectrum reactor was designed with a long fuel cycle, and high core exit temperature. These features are desirable in a reactor designed to provide process heat applications such as oil refinery needs of heat and electricity. The reactor uses the binary salt NaF-BeF₂ as the primary coolant, and uses U-Zr-H as fuel and to provide neutron moderation. Design options were studied under the constraints of maximum fuel temperature of 720[degree]C, and a vessel wall temperature similar to today's reactors (i.e. limited to 430[degree]C). The selected design achieves a core power density of 26.35 kW/1 and can achieve a 10 year core lifetime before refueling. The core has a pressure drop of 1.03 MPa, and uses a hexagonal lattice with small, 7mm outer diameter fuel pins. The pins are arranged in a tight 1.08 pitch-to-diameter ratio, and use wire wrap spacers and assembly ducts for mechanical support. Silicon carbide is used for cladding, and the fuel-clad gap is filled with a low-melting point liquid metal to act as a thermal bond between the fuel and the cladding. The core is encircled with two rows of reflector assemblies and one row of thermal shield assemblies. A layer of graphite provides thermal insulation to the reactor vessel. / (cont.) The primary coolant is coupled to a CO₂ power cycle via heat exchangers located above the core in the same vessel. The reactor vessel is 8.5m in diameter and 15.3m tall, which achieves the goal of a vessel that can be produced off-site and transported via barge, but not by truck or train. The intermediate heat exchanger is designed with 1 cm outer diameter tubing with internal helical ribs. The tubes are arranged in a square array with a pitch to diameter ratio of 1.2. With a core exit temperature of 570[degree]C, the supercritical CO₂ power conversion system achieves, according to previous studies at MIT, a net efficiency of 45.7%. A comparison is made to other integral medium reactors (i.e those with the heat exchanger and the core in the same vessel). This reactor has the advantage of low pressure and high thermal conversion ratio compared to the IRIS water cooled reactor. / by Joshua Whitman. / S.M.

Thesis (S.M. and S.B.)--Massachusetts Institute of Technology, Dept. of Nuclear Science and Engineering, 2009. / Cataloged from PDF version of thesis. / Includes bibliographical references (p. 123-126). / An experimental study was performed to determine thermal performance and adhesion effects of a combined nanoparticle and polymeric dispersion coating. The critical heat flux (CHF) values and nucleate boiling heat transfer coefficients (HTC) of nickel wires pre-coated using 1.0% alumina, 0.1% alumina, 500ppm polyallylamine hydrochloride (PAH), and 0.1% alumina combined with 500ppm PAH dispersions were determined using the pool-boiling method. The adhesion of 0.1% alumina and combined 0.1% alumina and 500ppm PAH coatings was evaluated using the tape and modified bend test methods. Results of the pool boiling experiments showed that the wire heaters pre-coated with combined 0.1% alumina and 500ppm PAH dispersion increase the CHF in water by -40% compared to bare wire heaters, compared to an enhancement of -37% with a 0.1% alumina coating. The combined 0.1% alumina and 500ppm PAH dispersion degrades the wire HTC by less than 1%, compared to a degradation of over 26% with a 0.1% alumina coating. Results from the tape test indicate qualitatively that the combined 0.1% alumina and 500ppm PAH dispersion coating adheres better than the 0.1% alumina nanoparticle coating. Results from the modified bend test showed that the combined 0.1% alumina and 500ppm PAH dispersion coating did not fail at the failure strain of the 0.1% alumina nanoparticle coating (8.108x 10-4). The addition of PAH to alumina nanofluid for creating a nanoparticle coating through boiling deposition was found to improve both coating thermal performance and adhesion over the pure alumina nanofluid. / by Bronwyn K. Edwards. / S.M.and S.B.

Thesis: S.M., Massachusetts Institute of Technology, Department of Nuclear Science and Engineering, 2014. / Cataloged from PDF version of thesis. / Includes bibliographical references (pages 105-112). / Fouling has been a persistent problem for fluid-solid interfaces, varying from heat exchangers to ship hulls. In nuclear power plants, buildup of CRUD can accelerate corrosion, increase pressure drops, cause axial power shifts, and increase radiation dose to workers. In this experiment, we studied the effect of induced microcavity and micropost patterns on CRUD morphology at atmospheric pressure. Samples with various pitches of microcavities and microposts were boiled in simulated PWR coolant. The effect of heat flux on CRUD formation was also examined. The experiments support previous theories that subcooled nucleate boiling is enhanced at microcavities, and that CRUD deposition is promoted at those sites. A thin ring of particles, less than a micrometer in width, was seen around most microcavities even when the surrounding areas were crudded. However, some of the samples with microcavity pitches 50 pm and smaller experienced a pattern of clean and crudded areas, with the region around the microcavities free of CRUD more often than not. A theory is put forth that this is caused by interfering forces due to bubbles collapsing. The micropost samples showed substantial CRUD growth, particularly for the sample subjected to a high heat flux. The results of this experiment could have major implications for development of a self-cleaning material for heated surfaces. / by Leigh Lin. / S.M.

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). / 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. / by Sarah M. Don. / S.M. / S.B.

Thesis: S.B., Massachusetts Institute of Technology, Department of Nuclear Science and Engineering, 2015. / Cataloged from PDF version of thesis. / Includes bibliographical references (pages 38-40). / This thesis presents the design and experimental validation of a photonic band gap (PBG) graded-index lens. PBGs can be used to achieve sub-wavelength focusing, which is not possible with lenses made from conventional dielectric materials. Subwavelength focusing is attractive for a variety of applications, including medicine, optics, and mining. By creating a focal spot smaller than the traditional diffraction-limited size, higher power density can be achieved in the focal spot. Further, subwavelength focusing is useful in imaging applications to view objects smaller than one wavelength. Using metamaterials is a common method for creating lenses that can beat the diffraction limit, and such devices have proven to be very successful. However, metamaterials are ill-suited for high-power microwave (HPM) applications because they can have very high electric fields in the metamaterial elements that can cause breakdown. PBG structures are capable of handling substantially higher power without facing problems with breakdown or arcing. The lens presented in this thesis is an attempt to create a PBG lens capable of sub-wavelength focusing specifically for use with HPM. Testing showed the lens achieved very good sub-wavelength focusing near the design frequency of 2.05 GHz, with focal spot widths between 0.58-0.75 times the traditional diffraction-limited size. The lens is capable of achieving subwavelength focusing over a range of frequencies of roughly 400 MHz, which is a additional advantage over low bandwidth metamaterial lenses. These results demonstrate significant progress in the development of novel electromagnetic materials suitable for high power applications. / by Samantha M. Lewis. / S.B.