Thesis (Sc. D.)--Massachusetts Institute of Technology, Dept. of Nuclear Science and Engineering, 2006. / Includes bibliographical references (p. 159-166). / In the first part of this thesis, the modes of microscopic energy fluctuations governing heat flow in nano-colloids are quantitatively assessed by combining linear response theory with molecular dynamics (MD) simulations. The intrinsic thermal conductivity is decomposed into self and cross correlations of the three modes that make up the microscopic heat flux vector, namely, the kinetic, the potential and the virial. By this decomposition analysis, the interplay between the molecular mechanisms that govern the variation of the thermal conductivity with volume fraction and solid-fluid interaction is examined. For a specific system of nanosized platinum clusters which interact strongly with host liquid xenon, a significant thermal conductivity enhancement is obtained as a result of self correlation in the potential energy flux. The effect saturates at higher volume fractions due to the cross-mode correlation between the potential and the virial flux. A strong solid-fluid coupling also introduces an amorphous-like structural transition and a pronounced cage effect that significantly reduces the self diffusion of the nano-clusters. These attendant structural and diffusive effects, unlike the self correlation of the potential flux, are amenable to experimental observations. The cluster-fluid interface is characterized by large fluctuations in the potential energy which is indicative of an unusual exchange of potential energy among the interfacial fluid atoms. For small nano-clusters, the interfacial layers interact with each other to form a percolating network. The research findings highlight the importance of surface interactions and show that the interfacial thermal resistance emanating from the self correlation of the collision flux is not the limiting mechanism for heat transfer in nano-colloids. / (cont.) This thesis also addresses several theoretical concerns regarding the microscopic thermal transport in colloids by using non-equilibrium molecular dynamics simulations (NEMD). The time averaged microscopic heat flux which assumes spatial homogeneity is shown to be applicable to nano-colloidal systems. Further, it is demonstrated that the thermal conductivity from a NEMD simulation is statistically equivalent to that of an equilibrium linear response evaluation only under certain dynamic conditions at the cluster-fluid interface. The concept of interfacial dynamical similarity is developed to establish this equivalence. The proposed thermal conduction model is consistent with several experimental observations such as the anomalous enhancement at small volume fractions with very small nanoparticles (3-10nm), limiting behavior at higher volume fractions, and the lack of correlation of the enhancement to the intrinsic thermal conductivity of the nano-clusters. The model also suggests possible avenues for optimizing the colloids by developing nano-clusters that have functionalized surface layers to maximize the interactions with the fluid atoms. In the second part of this thesis, smooth field estimators based on statistical inference and smoothing kernels are developed to transfer molecular data to the continuum for hybrid and equation-free multiscale simulations. The field estimators are then employed to implement coarse projection, a multiscale integration scheme, for a shear driven flow in an enclosure. This thesis shows that the spatial continuity and smoothness of the microscopically generated coarse variables, geometrically similar initial conditions and the separation of timescales are essential for the correct coarse field evolution with coarse projection. / by Jacob Eapen. / Sc.D.

Thesis (Ph. D.)--Massachusetts Institute of Technology, Dept. of Nuclear Science and Engineering, 2006. / Includes bibliographical references. / An investigation has been made into converting the MIT research reactor from using high enrichment uranium (HEU) to low enrichment uranium (LEU) with a newly developed fuel material. The LEU fuel introduces negative reactivity due to absorptions in 238U, which need to be compensated by higher initial content of 235U. Given that the new fuel density is much higher than the HEU density fuel, it is possible to obtain the necessary 235U content in the same core volume. A design of the MIT Nuclear Reactor is made using high density monolithic uranium-molybdenum fuel in an attempt to eliminate the reductions in neutron flux available to experiments due to the conversion to LEU fuel, as well as increasing the flexibility for meeting the needs of in-core experiments. The optimum configuration of fuel plates was made by varying the plate number and thicknesses and using a full-core model of the MITR for the Monte-Carlo transport code MCNP to determine the effect on flux and reactivity. In addition, the use of different moderator and fuel dummy materials as well as fixed absorbers was evaluated to optimize the neutron fluxes, reactivity and neutron spectrum available for experiments. The optimum reactor design consisted of the use of half-sized fuel elements made up of nine U-7Mo LEU fuel plates of 0.55 mm thickness with 0.25 mm finned aluminum cladding. This design also utilized solid beryllium dummies with boron fixed absorbers or solid lead dummies, depending on the in-core experiment flux and spectrum needs. Using MCODE, which links MCNP and the point depletion code ORIGEN, it was determined that the refueling interval of the LEU core would be about twice as long as the HEU core at the current power level of 5 MW. / (cont.) Thermal-hydraulic calculations using the multi-channel thermal-hydraulics analysis code MULCH-II indicated that the peak power channel will remain below the Onset of Nucleate Boiling under all normal operating conditions as well as loss of flow conditions. In addition, using MCNP and the thermal-hydraulics/point kinetics code PARET it was shown that all reactivity coefficients were negative and that the LEU core could withstand a step reactivity insertion of $3.69 without reaching cladding softening temperature, thus increasing the allowable reactivity for an incore experiment. Finally, it is possible to use the proposed design to increase the neutron flux by increasing core power, but with a correspondingly reduced refueling cycle length. / by Thomas Henderson Newton, Jr. / Ph.D.

Thesis: S.B., 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 418-420). / CRUD (Chalk River Unidentified Deposits) is buildup of metal oxides on the interior of nuclear reactors. This is caused by corrosion in reactor internals, leading to problems such as coolant contamination in porous deposits left by CRUD. CRUD has forced many nuclear reactors into temporary shutdown or production downgrades, costing millions of dollars US per reactor. If the CRUD growth factors could be fully understood, they could be controlled, and the CRUD problem could be eliminated altogether. Atomic force microscopy can be used to measure the force, or the strength of the CRUD-clad bond with different materials. This research focuses on answering this question: How does the force change between CRUD particles and different materials that could be used for reactor cladding? This study will analyze lab-grown CRUD samples on different substrate materials and characterize CRUD growth on each. It was found the CRUD-bond forces (from least to greatest) on silicon carbide (SiC), Titanium aluminum carbide (Ti2AlC), and max-phase zirconium alloy 211(Zr4M211) behaved similarly in air and in water. The forces on each surface increased with increasing dwell time for the Fe3O4 particle AFM tip; in contrast, most adhesion forces stayed constant with the NiO AFM tip. Furthermore, these CRUD forces were compared to other non-accident tolerant fuels, and there are cases in which non-ATF materials show more CRUD resistance (less adhesive force) than ATF-materials. This study's analysis could be applied to other materials to be used for reactor cladding. Once the material with the lowest-strength CRUD bond is identified and installed, the nuclear industry could save millions of dollars US per reactor fuel cycle. / by Rasheed Auguste. / S.B.

Thesis: S.M., Massachusetts Institute of Technology, Department of Nuclear Science and Engineering, 2015. / Cataloged from PDF version of thesis. / Includes bibliographical references (pages 72-75). / This work describes a framework built to simulate the thermal and hydrological processes in a deep borehole repository of spent nuclear fuel. Such simulation requires a fully coupled solver, capable of capturing the processes at the scale of tens of kilometers and millions of years. The MOOSE framework was chosen for this purpose, where the FALCON application, developed at INL, was adopted as baseline. This application had previously been applied for simulation of fluid flow and heat transport in geothermal reservoirs, and therefore provided a valuable reference. Additional features were implemented in FALCON in order to simulate deep borehole repositories. Solver options were adjusted for best performance. Code verification was performed on Rayleigh-Bnard convection in a porous medium. Cross-code validation was performed between the FALCON code and the FEHM code on a single borehole test case, and the thermal results were further compared to analytical and simplified numerical models, confirming the potential existence of a second peak of temperature at the scale of thousands of years. Two configurations for the borehole repository were analyzed. The first one consisted of an infinite array of boreholes, which allows one to significantly simplify the geometry, boundary conditions, and test code features. A parametric study of input parameters such as rock permeability, borehole spacing, and pitch length, was performed to assess thermal behavior of the repository. Analysis of the results led to the conclusion that the water flow in the caprock is driven mostly by thermal expansion of water. The displacement length of the water front was found to be negligible in comparison to the depth of the repository. The second configuration included a semi-infinite array of boreholes. This representation is a more realistic approximation of an actual repository, since it includes the modeling of the undisturbed rock surrounding the emplacement zone. It was shown that in this configuration convection can originate between the emplacement region and the rock outside the repository. At rock permeability higher than 10-16 m2 this mechanism can lead to an escape length of the water front larger than the burial depth. However, it was shown that the salinity gradient in the underground water can suppress convection and effectively eliminate water escape. / by Nazar Lubchenko. / S.M.

Thesis (Ph. D.)--Massachusetts Institute of Technology, Dept. of Nuclear Science and Engineering, 2008. / Includes bibliographical references (p. 81-87). / Coupled electron-nuclear spins are promising physical systems for quantum information processing: By combining the long coherence times of the nuclear spins with the ability to initialize, control, and measure the electron spin state, the favorable properties of each spin species are utilized. This thesis discusses a procedure to initialize these nuclear spin qubits, and presents a vision of how these systems could be used as the fundamental processing unit of a quantum computer. The focus of this thesis is on control of a system in which a single electron spin is coupled to N nuclear spins via resolvable anisotropic hyperfine (AHF) interactions. High-fidelity universal control of this le-Nn system is possible using only excitations on a single electron spin transition. This electron spin actuator control is implemented by using optimal control theory to find the modulation sequences that generate the desired unitary operations. Decoherence and the challenge of making useful qubits from these systems are also discussed. Experimental evidence of control using an electron spin actuator was acquired with a custom-built pulsed electron spin resonance spectrometer. Complex modulation sequences found by the GRadient Ascent Pulse Engineering (GRAPE) algorithm were used to perform electron spin echo envelope modulation (ESEEM) experiments and simple preparation-quantum operation-readout experiments on an ensemble of 1e-1n systems. The data provided evidence that we can generate any unitary operation on an AHF-coupled 1e-1n system while sitting on a single transmitter frequency. The data also guided design of the next iteration of these experiments, which will include an improved spectrometer, bandwidth-constrained GRAPE, and samples with larger Hilbert spaces. / by Jamie Chiaming Yang. / Ph.D.

Thesis (S.M.)--Massachusetts Institute of Technology, Dept. of Nuclear Science and Engineering, 2007. / Includes bibliographical references (p. 123-126). / The JAERI and NACOK experiments examine the combined effects of natural convection during an air ingress event: diffusion, onset of natural circulation, graphite oxidation and multicomponent chemical reactions. MIT has been benchmarking JAERI tests using the FLUENT code for approximately three years [1]. This work demonstrated that the three fundamental physical phenomena of diffusion, natural circulation and then chemical reactions can be effectively modeled using computational fluid dynamics. The latest series of tests conducted at the NACOK facility were two graphite corrosion experiments: The first test consisted of an open chimney configuration heated to 650C with a pebble bed zones of graphite pebbles and graphite reflectors. The second test is a similar test with a cold leg adjacent to the hot channel with an open return duct below the hot channel. Natural circulation, diffusion and graphite corrosion were studied for both tests. Using and adapting previous computational methods, the FLUENT code is used to blind benchmark these experiments. The objective is to assess the adequacy of the modeling method used in this blind bench-marking analysis by comparing these blind test predictions to the actual data and then modify the model to improve predictive capability. Ultimately, the objective is to develop a benchmarked analysis capability that can be used for real reactors calculations, and to improve the understanding of the physical phenomena taking place during an air ingress event. This thesis presents the modeling process of these experiments, the blind model results and the comparison of the blind computed data with experimental data. Sensitivity studies provide a good understanding of the different phenomena occurring during an air ingress event. The blind benchmarking demonstrates the ability of FLUENT to model satisfactorily in full scale the NACOK air ingress experiment. The blind models are then improved to successfully model air ingress events. An important finding of this work is that there is great variability in graphite corrosion data and that good qualification of specific graphite used is vital to predicting the effects of an air ingress event. / by Marie-Anne V. Brudieu. / S.M.

Thesis: S.B., Massachusetts Institute of Technology, Department of Nuclear Science and Engineering, 2018. / Cataloged from PDF version of thesis. / Includes bibliographical references (pages 41-43). / Latent heat energy storage (LHS) systems offer multiple advantages over sensible heat storage systems in terms of energy density, cost, and cycle efficiency. LHS systems use a phase change medium (PCM) that transfer heat at constant temperature. The use of metallic PCMs allows LHS to be used with high-temperature heat sources, such as the Fluoride-Salt-Cooled High- Temperature Reactor, and the use of a metallic heat transfer fluid (HTF) allows for a high rate of heat transfer. In this thesis, a 0.993 GWh thermal battery (ALTA) comprised of an aluminum PCM and sodium HTF is designed and evaluated in the steady-state. A novel rod design is also proposed to accommodate aluminum expansion as it changes phases. A final capital cost of $25.00/kWh is found, suggesting that ALTA is economically competitive. / by Kendall A. Fears. / S.B.

Thesis: S.M., Massachusetts Institute of Technology, Department of Nuclear Science and Engineering, 2016. / 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 64-68). / Limited data has been published (especially on experimental work) on integrated multilayer SiC/SiC prototypical fuel cladding. In this work the mechanical performance of three unique architectures of three-layer silicon carbide (SiC) composite cladding is experimentally investigated under conditions associated with the loss of coolant accident (LOCA), and analytically under various conditions. Specifically, this work investigates SiC cladding mechanical performance after exposure to 1,400°C steam for 48 hours and after thermal shock induced by quenching from 1,200°C into either 100°C or 90°C water. Mechanical performance characteristics are thereafter correlated with sample architecture through void characterization and ceramography. The series with a reduced thickness did not have a pseudo-ductile regime due to overloading of the composite layer. The presence of the axial tow did not yield significant difference in the mechanical behavior most likely because samples were tested in the hoop direction. While as-received and quenched samples behaved similarly (pseudo ductile failure except for one series), non-frangible brittle failure (single-crack failure with no release of debris) was systematically observed after oxidation due to silica buildup in the inner voids of the ceramic matrix composite (CMC) layer. Overall, thermal shock had limited influence on sample mechanical characteristics and oxidation resulted in the formation of silica on the inner wall of the CMC voids leading to the weakening of the monolith matrix and brittle fracture. Stress field in the cladding design is simulated by finite element analysis under service and shutdown conditions at both the core's middle height and at the end of the fuel rod. Stresses in the fuel region are driven by the thermal gradient that creates stresses predominantly from irradiation induced swelling. At the endplug, constraints are mainly mechanical. Stress calculations show high sensitivity to the data scatter and especially swelling and thermal conductivity. No cladding with the design studied here can survive either service or shutdown conditions because of the high irradiation-induced tensile stresses that develop in the hot inner monolith layer. It is shown that this peak tensile stress can be alleviated by adjusting the swelling level of the different layers. The addition of an under-swelling material such as PyC or Si can reduce the monolith tensile stress by 10%. With a composite that swells 10% less than the monolith, the stress is reduced by 20%. / by Pierre Guenoun. / S.M.

Thesis (S.M.)--Massachusetts Institute of Technology, Dept. of Nuclear Science and Engineering, 2009. / Page 208 blank. Cataloged from PDF version of thesis. / Includes bibliographical references (p. 159-160). / The supercritical carbon dioxide (SC0₂) power conversion system has been suggested for use with many of the Generation IV nuclear reactors. The SC0₂ cycle is highly attractive because of its low operating temperatures and high efficiency associated with working near the critical point of CO2. Unfortunately, the appealing features of using C0₂ near its critical point create complications in control. The Transient SC0₂ Cycles Code (TSCYCO) has been developed as a transient simulation control design and cycle scoping code for the recompression SC0₂ Brayton cycle. It is based off of the SC0₂ Power Systems (SCPS) code, and incorporates many improvements and modifications. Written in FORTRAN 90, TSCYCO uses a lumped parameter model and a momentum integral model approach. The code uses a semi-implicit solution process and implements Gaussian elimination to solve the system of equations. Transient behavior of the printed circuit heat exchangers is determined via the previously developed code HXMOD. Turbomachinery performance is modeled using the Real Gas Radial Compressor (RGRC) code with a scaling scheme for off-design conditions. Currently, TSCYCO has the capability of modeling several transients, including: loss of external load (LOEL), power load change, and cycle low-temperature change. Simulations show that TSCYCO can be run at quasi-steady state for an indefinite period of time. In the case of a 10% LOEL, the axial turbine experiences choke as a result of shaft overspeed. Turbine choke can be avoided if one bypasses more flow during LOEL. / (cont.) Moreover, one can incorporate more accurate axial turbine performance models to account for shaft speed variation. TSCYCO experiences instabilities when operated too closely to the critical point of C0₂. This could be remedied with a more robust Runge-Kutta solution method. / by Tri Q. Trinh. / S.M.

Thesis (Ph. D.)--Massachusetts Institute of Technology, Dept. of Nuclear Science and Engineering, 2012. / Cataloged from PDF version of thesis. / Includes bibliographical references (p. 225-237). / This thesis focuses on several aspects of the Lower Hybrid (LH) wave physics, the common theme being the development of full-wave simulation codes based on Finite Element Methods (FEM) used in support of experiments carried out on the Alcator C-Mod tokamak. In particular, two non-linear problems have been adressed: high power antenna-plasma coupling and current drive (CD). In both cases, direct solution of the wave equation allowed testing the validity of approximations which were historically done and consider full-wave effects and realistic geometries. The first code, named POND, takes into account the interaction of high power LH waves and the plasma edge based on the non-linear ponderomotive force theory. Simulations found the effect of ponderomotive forces to be compatible with the density depletion which is measured in front of the antenna in presence of high power LH waves. The second code, named LHEAF, solves the problem of LH wave propagation in a hot non- Maxwellian plasma. The electron Landau damping (ELD) effect was expressed as a convolution integral along the magnetic field lines and the resultant integro-differential Helmholtz equation was solved iteratively. A 3D Fokker-Planck code and a synthetic Hard X-Ray (HXR) diagnostic modules are used to calculate the self-consistent electron distribution function and evaluate the resulting CD and bremsstrahlung radiation. LHEAF has been used to investigate the anomalous degradation of LHCD efficiency at high density. Results show that while a small fraction of the launched power can be absorbed in the SOL by collisions, it is a strong upshift in the nii spectrum that makes the overall LHCD efficiency low by allowing the waves to Landau damp near the edge. Wavelet analysis of the full-wave fields identified spectral broadening to occur after the waves reflect and propagate in the SOL. This work explains why on Alcator C-Mod the eikonal approximation is valid only in the low to moderate density regime, and why parasitic phenomena introduced in previous work can reproduce phenomenologically well the experimental results. / by Orso Meneghini. / Ph.D.