Thesis (Ph. D.)--Massachusetts Institute of Technology, Dept. of Nuclear Engineering, 2005. / Includes bibliographical references (p. 119-124). / Solid state Nuclear Magnetic Resonance (NMR) allows us to explore a large coherent spin system and provides an ideal test-bed for studying strongly interacting multiple-spin system in a large Hilbert space. In this thesis, we experimentally investigate the spin dynamics in a rigid lattice of dipolarly coupled nuclear spins using multiple quantum NMR spectroscopy. Encoding multiple quantum coherences (MQC) in an arbitrary quantizing axis is developed. We utilized this method to encode coherence numbers in an orthogonal basis to Zeeman basis and showed that the dipolar-ordered state is a two spin correlated state, and confirmed the presence of the ... (flip-flop) terms in the experimentally prepared dipolar-ordered state. A new experimental investigation of the problem of the NM:R free induction decay (FID) in a lattice of spin 1/2 nuclei is presented to verify the multi-spin nature of the FID and the dominant role of the geometrical arrangement of the spins in the development of higher order correlations under the dipolar evolution. To study the dynamics and the controllability of these multiple spin correlations, effective decay times of individual coherence orders are measured under the dipolar interaction and under the control sequence that suppresses the dipolar evolution. / (cont.) It is seen that the decay time of each coherence order becomes shorter and more uniform among different coherence orders as the spin correlation size grows larger in both cases. Additional work has been done in this thesis, toward creating a pure state in solid state nuclear spins by transferring polarization from electron spins, i.e Dynamic Nuclear Polarization (DNP). A new cryogenic DNP probe was developed enabling multiple pulse irradiations at low temperature with enhanced polarization. / by HyungJoon Cho. / Ph.D.

Thesis (S.M.)--Massachusetts Institute of Technology, Dept. of Nuclear Engineering, 2005. / Includes bibliographical references (p. 135-137). / The first portion of this work is a comprehensive analysis of the chemical environment in a High Temperature Gas-Cooled Reactor TRISO fuel particle. Fission product inventory versus burnup is calculated. Based on those results a thermodynamic analysis is performed to determine fission product vapor pressures, oxygen partial pressure, and carbon monoxide and carbon dioxide gas pressures within the fuel particle. Using the insight gained from the chemical analysis, a chemical failure model is incorporated into the MIT fuel performance code, TIMCOAT. Palladium penetration of the SiC layer is added to the fracture mechanics failure model. Rare-earth fission product and palladium corrosion of the SiC layer are additionally modeled. The amoeba effect is added as a new failure mode. The palladium penetration model has the most significant result on the overall fuel performance model and increases the number of predicted particle failures. The thinning of the SiC layer due to fission product corrosion has a slight effect on the overall fuel performance model. Finally, the amoeba effect model does not lead to any particle failures, but adds to the completeness of the overall model. / by Jane T. Diecker. / S.M.

Thesis (S.M.)--Massachusetts Institute of Technology, Dept. of Nuclear Engineering, 2005. / Includes bibliographical references. / The goal of this thesis is to design an experimental thermal-hydraulic loop capable of generating accurate, reliable data in various convection heat transfer regimes for use in the formulation of a comprehensive convection heat transfer correlation. The initial focus of the design is to ensure that the loop will be able to generate the convection flow regimes found in post Loss of Coolant Accident (LOCA) operation of a Gas-cooled Fast Reactor (GFR). As a result a scaling analysis of the proposed test facility was undertaken to demonstrate that the proposed loop would be able to operate in these aforementioned regimes. Having verified that the experimental loop could operate in the regimes of interest the next stage in the project was construction of the loop. Following construction of the loop and necessary instrumentation, an uncertainty analysis of the facility was conducted with the goal of determining the uncertainty associated with the calculation of heat transfer coefficients from the experimental data. The initial results were discouraging as the uncertainty calculated was large, ranging from -10-60%. After performing a heat transfer coefficient uncertainty analysis, we observed that the bulk of the uncertainty was clue to heat loss from the fluid to the environment. / (cont.) Therefore, guard heaters were implemented into the loop design, to match the inner surface temperature of the insulation to the wall temperature of the test section, which allows minimization of heat loss to about zero. This resulted in the considerable reduction in heat transfer coefficient uncertainty to -8-15%. / by Peter A. Cochran. / S.M.

Thesis (Ph. D.)--Massachusetts Institute of Technology, Dept. of Nuclear Engineering, 2005. / Includes bibliographical references (p. 245-248). / In order to fulfill the goals set forth by the Generation IV International Forum, the current NERI funded research has focused on the design of a Gas-cooled Fast Reactor (GFR) operating in a Breed and Burnm (B&B) fuel cycle mode. B&B refers to a once-through fuel cycle where low enriched uranium (less than 5 w/o 235U in U) subcritical assemblies are loaded into the core in equilibrium, yet in-situ plutonium breeding carries the fuel through a discharge burnup on the order of 150 MWD/kgHM. The B&B fuel cycle meets the GenIV goals of sustainability, economics, and proliferation resistance by increasing fuel burnup without the need for spent fuel reprocessing, recycle, or reuse of any kind. The neutronic requirements for B&B are strict and require an ultra-hard neutron spectrum. Therefore, the GFR is ideally suited for this fuel cycle. In the present work the B&B GFR concept evolved into two practical reactor designs, both of which build on extensive previous gas-cooled reactor design experience. The first version is the "demonstration" concept using highly neutronically reactive U15N fuel in a hexagonal pin fuel array that is nearly 50 v/o fuel. The core is helium cooled, with an outlet temperature of 570 °C. / The helium primary circuit is coupled to a steam Rankine power conversion system essentially identical to that for the British Advanced Gas-cooled Reactors. One advantage of the low coolant temperature compared to other GenIV GFR concepts is that it allows for the use of oxide dispersion strengthened stainless steels (ODS) in core. The fuel is manufactured using advanced vibration compaction techniques, clad in ODS, and vented in order to achieve the high burnup goal. The second version, the "advanced" concept builds on the experience of the demonstration concept to develop a B&B GFR without the need for expensive U'5N fuel. In order to substitute the nitride fuel with carbide, significantly higher heavy metal loadings are required (60 v/o fuel for UC versus 50 v/o fuel for U'5N) which are not practically achievable with a conventional pin fuel array. Therefore, an innovative tube-in-duct assembly design was proposed to achieve B&B operation with the less neutronically reactive carbide fuel. The advanced core offers significantly reduced natural uranium requirements and lower equilibrium fuel cycle costs (5 mills/kWhre) compared with conventional light water reactors (7 mills/kWhre), as the burnup is tripled for the same reload enrichment. / (cont.) The B&B GFR designs, though requiring active decay heat removal, are semi-self-regulating from a reactivity feedback standpoint and are designed to withstand all plausible accident scenarios, including loss of flow, loss of heat sink, and transient overpower all without scram. Reactor pressure vessel blowdown (LOCA) was investigated and while the B&B GFR has a low positive coolant void reactivity (less than 1$), the added reactivity during blowdown is compensated through other strong negative reactivity feedback mechanisms, thereby allowing for the safe operation of the B&B GFR. / by Peter Yarsky. / Ph.D.

Thesis (Ph.D.)--Massachusetts Institute of Technology, Dept. of Nuclear Engineering, 1983. / MICROFICHE COPY AVAILABLE IN ARCHIVES AND SCIENCE / Bibliography: leaves 135-139. / by Reza Najafabadi Farahani. / Ph.D.

Thesis (Ph.D.)--Massachusetts Institute of Technology, Dept. of Nuclear Engineering, 1984. / MICROFICHE COPY AVAILABLE IN ARCHIVES AND SCIENCE. / Includes bibliographical references. / by Kie L SooHoo. / Ph.D.

Thesis (M.S.)--Massachusetts Institute of Technology, Dept. of Nuclear Engineering, 1984. / MICROFICHE COPY AVAILABLE IN ARCHIVES AND SCIENCE. / Includes bibliographical references. / by Susan M. Reilly. / M.S.

Thesis (Ph. D.)--Massachusetts Institute of Technology, Dept. of Nuclear Engineering, 2001. / Includes bibliographical references. / The objective of this thesis is to clarify the ways that organizational factors influence nuclear power plant performance in order to improve performance. Therefore, this thesis studied the nuclear power plant organizational environment by identifying and detailing its important work processes. These work processes are: the Work Request Work Process; the Condition Reporting Work Process; the External Operating Experience Work Process; the Design Change Process; and the Procedure Change Work Process. Using this information, a methodology of incident investigation that targets organizational deficiencies contributing to events was developed. Using this methodology to analyze recent significant incidents, a list of important organizational factors and the context within which they influence the successful completion of tasks was identified. These factors and the context within which they influence performance are: 1) Communication-Pervasive-Most important between different units and departments; 2) Formalization-Execution; 3) Goal Prioritization-Prioritization; 4) Problem Identification-Planning, scheduling, and return to normal line-up; 5) Roles and Responsibilities-Execution; and 6) Technical Knowledge (job specific knowledge and broad based knowledge)-Job specific knowledge-execution/Broad based knowledge -prioritization, planning, scheduling, and other tasks. Although safety culture and organizational learning are not listed, they are important. / (cont.) The reason for their exclusion was that they are not single organizational factors useful when cited in incident investigations. Rather, safety culture is a term used to describe all organizational factors, including organizational structure that impact performance. Similarly, organizational learning was excluded because it is a collection of programs, processes, individual attitudes and culture responsible for learning. Although organizational learning was not listed, it was studied resulting in the development of the Utilization of Operating Experience Work Process. The Utilization of Operating Experience Work Process consists of the following seven steps: 1) Identification; 2) Screening/Prioritization/Dissemination; 3) Investigation/Evaluation; 4) Development; 5) Implementation; 6) Closeout; and 7) Verification/Validation. Since prioritization was identified as important in the above work process and the analysis of significant events, a methodology for the prioritization of work activities at nuclear power plants was developed. This methodology produces a prioritization tool that assigns a numerical performance index to each item requiring prioritization is developed. Applying the methodology at Seabrook Station produced a tool that allowed those who prioritize external operating experience to more efficiently and accurately prioritize. In addition to the success of the application at Seabrook, a workshop held at MIT with experts in prioritizing external operating experience who further validated the methodology and the resulting tool. / by Richard Scott Weil. / Ph.D.

Thesis (Ph. D.)--Massachusetts Institute of Technology, Dept. of Nuclear Engineering, 2001. / Includes bibliographical references (p. 140-143). / Fast Neutron Resonance Radiography (NRR) has been devised as an elemental imaging method, with immediate applications to detecting explosives and drugs in passenger suitcases. In the NRR method, the 2-D elemental mapping of hydrogen, carbon, nitrogen, oxygen and the sum of other elements are obtained from fast neutron radiographic images taken at different neutron energies chosen to cover the resonance cross section features of one or more elements. A radiographic image provides the 2-D mapping of the sum of elemental contents (weighted by the attenuation coefficients) and transmission measurements taken at different neutron energies form a set of linear equations, which can be solved to map individual elemental contents. Explosives and drugs can be identified by their characteristic elemental composition. The object-detector assembly rotates around the neutron source and different energy (2-6 MeV) neutrons can be obtained at different angles from a DD neutron source. A fixed-energy RFQ (2.3 MeV deuteron energy) with a thick target (0.4-0.8 MeV) can be used to generate fast neutrons. The physics of image formation in fast neutron radiography has been studied and a thick (4 cm) plastic scintillator has been recommended as the neutron detector. / (cont.) Variance-checking median filters have been recommended for removing sparkles in the digital image and the direct Householder Transform has been used to solve the ill-conditioned Least-Squares problem. Source shielding for 2x109 neutrons/second has been suggested. Experiments and simulation have proven NRR to be effective. Explosives can be identified with their high nitrogen and oxygen content and drugs with their high C/O ratio. With a 10 [mu]A deuteron current and 6 atmosphere-cm D2 gas target, total imaging time for each piece is expected to be 1-2 minutes. Because multiple objects can be inspected at the same time, it is possible to meet the FAA standard of 8 seconds per piece. Spatial resolution of the image is estimated to be -7.5 mm. / by Gongyin Chen. / Ph.D.

Thesis (Ph. D.)--Massachusetts Institute of Technology, Dept. of Nuclear Engineering, 2001. / Vita. / Includes bibliographical references (p. 297-320). / Dislocation is a line defect in crystalline materials, and a microscopic carrier of plastic deformation. Because dislocation has both a localized core and a long-range stress field, linking atomistic and meso scales is often the most challenging step in studying its dynamics. This Thesis presents theories and simulations of dislocations in Si and BCC transition metals, with emphasis on the atomistic-mesoscale coupling. Contributions are made in both methods development and mechanistic understanding of dislocation mobility. For atomistic studies of defects embedded in a mesoscale surrounding, we have given rigorous treatments of two types of boundary effects. A method is derived for quantifying artificial image energies in dislocation simulations with a periodic cell, in which a longstanding conditional convergence problem in lattice summation is resolved. We have also developed a systematic approach based on the linear response theory, which minimizes boundary wave reflections in molecular dynamics simulations without artificial damping. When predictive models are confronted with experiments at the level of mesoscale kinetics, the challenge is to properly incorporate atomistic details into a coarse-grained simulation. / (cont.) We have investigated dislocation core and kink mechanisms and obtained deeper understandings on the shuffle-glide controversy in Si and edge versus screw dislocations in BCC Mo, with some of these breakthroughs related to a better control of artificial boundary effects. The atomistic-mesoscale coupling is then manifested in our formulation of a kinetic Monte Carlo description of dislocation glide in Si at the mesoscale, based on kink mechanisms. As a result, the nature of "weak obstacles" to kink propagation, a long-standing postulate for interpreting low stress dislocation mobility data, is clarified. This model is then generalized to incorporate cross slip for modeling screw dislocation motion in a BCC lattice. Lastly, a physically-motivated procedure is derived for removing the stress singularity in mesoscale dislocation dynamics simulations. / by Wei Cai. / Ph.D.