Thesis: Ph. D., 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 136-140). / Monte Carlo is increasingly being used to perform high-fidelity, steady-state neutronics analysis of power reactor geometries on today's leadership class supercomputers. Extending Monte Carlo to time dependent problems has proven to be a formidable challenge due to the significant computational resource and data processing requirements. In this thesis, a transient methodology is proposed and implemented to enable accurate and computationally tractable time dependent Monte Carlo analysis. The frequency transform method has been described and implemented in Monte Carlo for the first time. The attractiveness of this method lies in its ability to accurately capture the space and time dependent distribution of the delayed neutron source throughout a transient. Nuances to the algorithmic implementation are described and validated through a series of simple analytical test problems. Comparison with the adiabatic method currently employed for Monte Carlo transient analysis shows significant improvement in the spatial distribution and magnitude of the power for a negative reactivity insertion transient in the 2D and 3D C5G7 geometry. To aid in understanding the effect of statistical uncertainty in the tallied quantities on the time dependent flux solution, a simplified point kinetics model was developed and used for insightful analysis on simple transient test problems. This revealed how the time dependent flux profiles for a series of independent trials can be approximated by a normal distribution at low uncertainties in the tallied reactivity, but deviates from a normal distribution at relatively modest uncertainties in reactivity. Given the compuational constraints of solving large problems, having a simple model that can provide insight on the expected behavior and flux distribution can be very valuable. The frequency transform methodology belongs to a class of indirect space-time factorization methods that perform high-order calculations (e.g. Monte Carlo) over long time steps and low-order, computationally-efficient calculations (e.g. Point Kinetics) over short time steps as an approach to balance performance and accuracy. The coarse mesh finite difference (CMFD) diffusion operator is employed as the low-order solver in Monte Carlo transient analysis for the first time. The CMFD diffusion operator is attractive due to its potential to increase the time step size between the computationally expensive high-order solves. Implementing this methodology is important as continuous energy Monte Carlo is reactor-agnostic and able to treat complex geometries without difficulty, opening up the possibility of solving transients on new experimental geometries for which there is little data. / by Samuel Christopher Shaner. / Ph. D.

Thesis (Ph. D.)--Massachusetts Institute of Technology, Dept. of Nuclear Science and Engineering, 2006. / Includes bibliographical references (p. 133-142). / Coherence and entanglement in multi-spin systems are valuable resources for quantum information processing. In this thesis, I explore the manipulation of quantum information in complex multi-spin systems, with particular reference to Nuclear Magnetic Resonance implementations. In systems with a few spins, such as molecules in the liquid phase, the use of multi-spin coherent states provides a hedge against the noise, via the encoding of information in logical degrees of freedom distributed over several spins. Manipulating multi-spin coherent states also increases the complexity of quantum operations required in a quantum processor. Here I present schemes to mitigate this problem, both in the state initialization, with particular attention to bulk ensemble quantum information processing, and in the coherent control and gate implementations. In the many-body limit provided by nuclear spins in single crystals, the limitations in the available control increase the complexity of manipulating the system; also, the equations of motion are no longer exactly solvable even in the closed-system limit. Entanglement and multi-spin coherences are essential for extending the control and the accessible information on the system. I employ entanglement in a large ensemble of spins in order to obtain an amplification of the small perturbation created by a single spin on the spin ensemble, in a scheme for the measurement of a single nuclear spin state. I furthermore use multiple quantum coherences in mixed multi-spin states as a tool to explore many-body behavior of linear chain of spins, showing their ability to perform quantum information processing tasks such as simulations and transport of information. / (cont.) The theoretical and experimental results of this thesis suggest that although coherent multi-spin states are particularly fragile and complex to control they could make possible the execution of quantum information processing tasks that have no classical counterparts. / by Paola Cappellaro. / Ph.D.

Thesis: Ph. D., Massachusetts Institute of Technology, Department of Nuclear Science and Engineering, 2013. / 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. / Disruptions, the sudden termination of tokamak fusion plasmas by instabilities, have the potential to cause severe material wall damage to large tokamaks like ITER.e mitigation of disruption damage is an essential part of any fusion reactor system. Massive gas injection (MGI) rapid shutdown is a technique in which large amounts of noble gas are injected into the plasma in order to safely radiate the plasma energy evenly over the entire plasma-facing first wall. However, it has been observed that this energy is not radiated evenly: it can have significant asymmetries, which could cause melting in large devices even in the case of a successful rapid shutdown. The first rapid shutdown experiments using multiple gas injectors on any tokamak were conducted on Alcator C-Mod. A dedicated toroidal array of fast ultraviolet photodiodes was installed in order to diagnose toroidal radiation asymmetries during the thermal quench (TQ). It is found that the radiation asymmetry is controlled by a low-n brightness mode in the TQ phase of rapid shutdowns. is mode sometimes rotates, and the rate of rotation sets the integrated radiation asymmetry in the TQ. It is proposed that this brightness feature is caused by the transport of energy from the hot plasma core to the radiative edge by the MHD flow at one phase of an n = 1 global MHD mode. is phenomenology is confirmed by extended MHD simulation using the NIMROD code. An exponentially growing n = 1 magnetic mode is observed during the pre-TQ phase of MGI rapid shutdowns; the saturation of this mode marks the beginning of the thermal quench. It is proposed that this mode is a magnetic island caused by a radiative tearing mode; the predicted growth rate is compared to the predictions of analytic theory. It is proposed that this mode is a magnetic island then couples to other global n = 1 MHD modes, causing the energy transport during the TQ. An important implication of this result is that simply adding more gas injectors cannot guarantee a symmetric rapid shutdown: the asymmetry is controlled by the behavior of the core MHD activity during the TQ. the implications of this rotating radiation asymmetry during the TQ of MGI rapid shutdown for the beryllium wall of ITER are discussed. / by Geoffrey Michael Olynyk. / Ph. D.

Thesis (Ph. D.)--Massachusetts Institute of Technology, Dept. of Nuclear Science and Engineering, 2011. / Cataloged from PDF version of thesis. / Includes bibliographical references (p. 148-156). / Nanofluids, engineered colloidal dispersions of nanoparticles in fluid, have been shown to enhance pool and flow boiling CHF. The CHF enhancement was due to nanoparticle deposited on the heater surface, which was verified in pool boiling. However, no such work has been done for flow boiling. Using a cylindrical tube pre-coated with Alumina nanoparticles coated via boiling induced deposition, CHF of water was found to enhance up to 40% compared to that of the bare tube. This confirms that nanoparticles on the surface is responsible for CHF enhancement for flow boiling. However, existing theories failed to predict the CHF enhancement and the exact surface parameters attributed to the enhancement cannot be determined. Surface modifications to enhance critical heat flux (CHF) and Leidenfrost point (LFP) have been shown successful in previous studies. However, the enhancement mechanisms are not well understood, partly due to many surface parameters being altered at the same time, as in the case for nanofluids. Therefore, the remaining objective of this work is to evaluate separate surface effect on different boiling heat transfer phenomena. In the second part of this study, surface roughness, wettability and nanoporosity were altered one by one and respective effect on quenching LFP with water droplet was determined. Increase in surface roughness and wettability enhanced LFP; however, nanoporosity was most effective in raising LFP, almost up to 100°C. The combination of the micro posts and nanoporous coating layer proved optimal. The nanoporous layer destabilizes the vapor film via heterogeneous bubble nucleation, and the micro posts provides intermittent liquid-surface contacts; both mechanisms increase LFP. In the last part, separate effect of nanoporosity and surface roughness on pool boiling CHF of a well-wetting fluid, FC-72, was investigated. Nanoporosity or surface roughness alone had no effect on pool boiling CHF of FC-72. Data obtained in the literature mostly for microporous coatings showed CHF enhancement for well wetting fluids, and existing CHF models are unable to predict the enhancement. / by Bao Hoai Truong. / Ph.D.

Thesis: Ph. D., Massachusetts Institute of Technology, Department of Nuclear Science and Engineering, 2014. / Cataloged from PDF version of thesis. / Includes bibliographical references (pages 191-193). / A new methodology has been developed that couples differential cross section data evaluation with integral benchmark analysis for improved uncertainty quantification. The new methodology was applied to the two new copper evaluations and resulted in an improved evaluation with smaller covariance data. Copper is a structural material in many nuclear applications, particularly those dealing with criticality safety. The current standard for the resonance evaluation of the two copper isotopes, ⁶³,⁶⁵Cu, has been determined to result in poor modeling performance. Therefore a new resonance evaluation of the two copper isotopes is vital to nuclear criticality safety applications. Performing a new resolved resonance region evaluation for copper has served as a backdrop to this work on developing new techniques for resolved resonance region evaluation. For the new evaluations, experimental cross section measurements have been carried out in the thermal energy region where no experimental data had previously been measured. Along the way, an automated routine was developed to aid with the determination of the quantum angular momentum of newly identified resonances. The impact of differential scattering cross sections with respect to angle was determined in the benchmarking process. The implications of the study of the impact of differential cross sections on criticality suggest a necessity for detailed treatment of the angular distributions during the evaluation process, as well as temperature broadening of the angular distributions for simulation applications. The formalism for temperature broadening of angular distributions has been derived and tested. The new evaluations were compared against the current ENDF/B-VII.1 standard on a set of 23 criticality safety benchmark models and displayed improved performance. In the new methodology developed for coupling of the differential and integral data evaluation, resonance parameters are directly and systematically adjusted based on feedback from integral benchmark experiments. Coupling this feedback directly to the resonance parameters gives the new method the advantage of implicitly adjusting all of the cross sections simultaneously, including the double differential cross sections. Based on integral feedback, the new methodology provides a way of updating the reported covariance of the resolved resonance region to reflect true state of knowledge. / by Vladimir Sobes. / 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. 149-152). / In order to provide improved predictions of Pellet Cladding Mechanical Interaction (PCMI) for the FRAPCON nuclear fuel performance code, a new model, the FRAPCON Radial-Axial Soft Pellet (FRASP) model, was developed. This new model uses 1.5D structural mechanics to represent both the fuel pellet and cladding along with their interaction via interfacial forces. The fuel pellet and cladding are modeled as concentric annular cylinders using similar governing equations with slight differences to allow for cracking of the semi-brittle fuel matrix and plastic behavior in a ductile cladding. By accounting for the structural mechanics of the fuel pellet, FRASP allows for stress-induced deformations which were previously unattainable with the rigid pellet model used by FRAPCON. Because of the significant differences between FRAPCON's previous mechanical model, FRACASI, and FRASP, simply replacing the treatment of PCMI within the code was not a viable option. This led to a complete replacement of FRACAS-I and all associated fuel rod structural calculations. Feedback effects are likely to result from such a major change due to the complexity of nuclear fuel simulation. The potential for these feedback effects dictated a preliminary validation of FRASP against FRACAS-I for typical case. This evaluation was not limited to the investigation of mechanical parameters, but covered a wide variety of predicted parameters by the new and unaltered versions of FRAPCON. The differences which were found in this validation were limited in nature and easily attributable to the differing assumptions of FRASP and FRACAS-I. The newly developed mechanical model was used with the improved fuel behavior models of FRAPCON-EP (Enhanced Performance) to assess the mechanical behavior of fuel rods with a composite silicon carbide (SiC) cladding under Pressurized Water Reactor (PWR) conditions. The fuel rod designs were selected to match previously chosen values for both solid and annular fuel pellets under current and uprated power conditions. Unlike FRACAS-I, which is hindered by the rigid pellet model, FRASP was able to successfully analyze PCMI behavior with the more rigid SiC, even though "hard contact" of the fuel and cladding was encountered. Simulations using the improved models showed that the SiC clad fuel rods may not provide adequate safety margins at the desired burnup, or simply fail to achieve their desired final burnup. Previous analyses which relied on FRAPCON-3.3 may have been overly optimistic in this regard. The new, more conservative predictions are largely due to FRASP's treatment of the inner radius of the annular fuel pellets, which was assumed not to change in previous versions of FRAPCON. These new findings suggest that SiC fuel rod general design and operation require further optimization. / by Alexander James Mieloszyk. / S.M.

Thesis: Ph. D., Massachusetts Institute of Technology, Department of Nuclear Science and Engineering, 2016. / Cataloged from student-submitted PDF version of thesis. / Includes bibliographical references (pages 175-184). / The objective of this thesis is to uncover the chemical states and atomic structure of the initial oxide on zirconium and the oxygen transport kinetics through this oxide under electric field. This goal is important for enabling more accurate zirconium oxidation models, for example for nuclear reactor materials, as well as for assessing the mechanisms that govern the performance of zirconia based technologies, such as redox based resistive switching memory devices, gate dielectric for metal oxide semiconductor devices, and electrolytes for solid oxide fuel cells ... / by Wen Ma. / Ph. D.

Thesis: Ph. D., 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 408-431). / According to the future of the nuclear fuel cycle study at MIT, a reactor with a conversion ratio around one can achieve desired objectives in the long-term sustainability of uranium and reduction of transuranic wastes. This finding relaxes the need for sodium fast reactors (SFR) in a closed-loop nuclear fuel cycle and enables high-conversion light water reactors (HC-LWR) to be used as an alternative. HC-LWRs have two major advantages over SFRs. First, apart from the reactor core, the remaining reactor system can be based on existing LWR technology. Second, extensive operating experience and a proven record of high reliability of LWRs would ease licensing and commercialization processes. Therefore, operating HC-LWRs instead of SFRs may be more economically and technically viable with lower capital and development cost for the near term. This type of reactor is being developed by Hitachi Ltd. under the name of resource-renewable boiling water reactor (RBWR). This study focuses on RBWR-TB2, transuranic burning version of RBWR. To demonstrate that the RBWR-TB2 can operate safely within design constraints and regulatory limits, the thermomechanical behavior of this reactor has been analyzed through fuel performance modeling. Due to its unique design characteristics, several physical phenomena at high temperature and high burnup typically ignored in most LWR fuel performance codes can potentially become active under RBWR's operating conditions. These phenomena involve migration of fuel constituents and fission products, the evolution of O/M ratio with burnup, high burnup structure (HBS) formation, accelerated corrosion, hot pressing, gaseous fuel swelling, hydride precipitation and hydrogen migration in the cladding. Semi-empirical models describing porosity and cesium migration behaviors have been replaced with mechanistic models. All of these phenomena have been successfully implemented in a modified version of FRAPCON-3.5 known as FRAPCON-3.5 EP where EP stands for enhanced performance. The fuel performance comparison between RBWR-TB2 and ABWR fuel rods suggest that because of high axial peaking factors and relatively flat power history, fuel temperature is significantly higher in fissile zones of the RBWR-TB2 leading to various undesirable effects such as excessive fission gas release and cladding deformation. Local fuel burnup in fissile zones of RBWR-TB2 is multiple times higher than that of ABWR leading to excessive fuel swelling, accelerated cladding oxidation, and PCMI at fissile-blanket interfaces. Even if the RBWR-TB2 has to operate under such demanding conditions with a small margin to fuel melting, a steady-state fuel performance analysis still shows that this reactor can operate safely with an acceptable thermo-mechanical performance. In the future optimization of RBWR-TB2 performance, several fuel design strategies are recommended based on a series of sensitivity studies. The sensitivity study on key design parameters indicates that using annular fuel geometry and more hypostoichiometric fuel (lower O/M ratio) could reduce fuel temperature at high burnup. For better resistance to cladding corrosion and PCMI, it is recommended to increase cladding thickness and decrease fuel density. / by Yanin Sukjai. / Ph. D.

Thesis (S.M.)--Massachusetts Institute of Technology, Dept. of Nuclear Science and Engineering, 2007. / Includes bibliographical references (leaves 84-85). / The United States Nuclear Regulatory Commission is responsible for the safe operation of the United States nuclear power plant fleet, and human reliability analysis forms an important portion of the probabilistic risk assessment that demonstrates the safety of sites. Treatment of post-initiating event human error probabilities by three human reliability analysis methods are compared to determine the strengths and weaknesses of the methodologies and to identify how they may be best used. A Technique for Human Event Analysis (ATHEANA) has a unique approach because it searches and screens for deviation scenarios in addition to the nominal failure cases that most methodologies concentrate on. The quantification method of ATHEANA also differs from most methods because the quantification is dependent on expert elicitation to produce data instead of relying on a database or set of nominal values. The Standardized Plant Analysis Risk Human Reliability Analysis (SPAR-H) method uses eight performance shaping factors to modify nominal values in order to represent the quantification of the specifics of a situation. The Electric Power Research Institute Human Reliability Analysis Calculator is a software package that uses a combination of five methods to calculate human error probabilities. Each model is explained before comparing aspects such as the scope, treatment of time available, performance shaping factors, recovery and documentation. Recommendations for future work include creating a database of values based on the nuclear data and emphasizing the documentation of human reliability analysis methods in the future to improve traceability of the process. / by Phillip E. Dawson. / S.M.

Thesis (S.M. and S.B.)--Massachusetts Institute of Technology, Dept. of Nuclear Science and Engineering, 2009. / 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. 123-130). / Surface effects on pool boiling heat transfer and the critical heat flux are well documented but poorly understood. This study investigates the pool boiling characteristics of various fluids, and demonstrates that surface effects can drastically alter the nucleate boiling heat transfer coefficient as well as the critical heat flux. Changes in surface morphology and surface chemistry are suspected to be the primary factors influencing pool boiling heat transfer. The relative impact of surface properties is shown to depend strongly upon the working fluid. To evaluate the effects of chemical constituency and surface texture on the pool boiling of water, nanoparticle thin-film coatings are applied to nickel and stainless steel substrates using the layer-by-layer assembly method. This study shows that such coatings, with thicknesses on the order of one micron or less, are capable of enhancing the critical heat flux of water up to 100%, and enhancing the nucleate boiling heat transfer coefficient over 100%. Through the use of thin-film coatings, the importance of nanoscale surface texture, porosity, and chemical constituency on boiling mechanisms is revealed. Low surface tension dielectric fluids, including a recently developed fluorinated ketone with a low global warming potential, are tested to determine their pool boiling heat transfer capabilities. The potential for nanoparticle-based pool boiling enhancement in well-wetting dielectric fluids is investigated. The role of surface wettability and adhesion tension on the incipience of boiling, nucleate boiling, and critical heat flux are considered. / (cont.) Results indicate that the low global warming potential fluorinated ketone may be a viable alternative in the cooling of electronic devices. Additionally, results demonstrate that enhancement of boiling heat transfer is possible for well-wetting dielectric fluids, with 40% enhancement in the critical heat flux using dilute suspensions of aluminum or silica nanoparticles in the fluorinated ketone. / by Eric Christopher Forrest. / S.M.and S.B.