Thesis (Ph. D.)--Massachusetts Institute of Technology, Dept. of Nuclear Engineering, 2003. / Includes bibliographical references (p. 127-133). / This electronic version was submitted by the student author. The certified thesis is available in the Institute Archives and Special Collections. / In this thesis I study control of quantum systems while implementing complex quantum operations. Through experimental implementations of such operations, I test the accuracy of control and provide methods for identifying the type and strength of experimental errors. The centerpiece of this work is the quantum Fourier transform (QFT), an essential gate for quantum algorithms and quantum simulations. Experiments are performed on a three qubit liquid-state nuclear magnetic resonance quantum information processor, and demonstrate salient features of the QFT in both of these venues. The first experiment exhibits the ability of the QFT to extract periodicity, a necessary process for many quantum algorithms. As an example of a quantum simulation, I implement a three qubit quantum baker's map, which is composed of QFTs, and discuss how various conjectures of quantum chaos could be experimentally realized on a quantum computer. Another example of complex quantum operations are 'pseudo-random' maps. These are operators which pass statistical tests of randomness but can be efficiently implemented on a quantum computer. I explore the importance of pseudo-random maps for the study of quantum chaos and a host of quantum information processing protocols. I also implement a set of such maps experimentally. In order to determine the type and strength of the errors effecting our implemetations, quantum process tomography is done on the QFT. / (cont.) From the constructed QFT superoperator and Kraus forms I show how best to analyze the data in order to extract information about coherent, incoherent, and decoherent errors. Finally, I explore fidelity decay as a signature of quantum chaos. The simulations performed concentrate on the exact determination of fidelity decay behavior for quantum chaotic systems, and attempt to identify properties of the evolution operator that cause the observed fidelity decay behavior. / by Yaakov Shmuel Weinstein. / Ph.D.

Thesis (Nucl. E. and S.M.)--Massachusetts Institute of Technology, Dept. of Nuclear Engineering, 2003. / Includes bibliographical references (p. 117-121). / This electronic version was submitted by the student author. The certified thesis is available in the Institute Archives and Special Collections. / A thermal hydraulic analysis of the International Reactor Innovative and Secure (IRIS) core has been performed. Thermal margins for steady state and a selection of Loss Of Flow Accidents have been assessed using three methodologies to account for uncertainty. The thermal hydraulic analysis has shown that the IRIS is designed with adequate thermal margin for steady state operation, the locked rotor/shaft shear accident (LR/SS) and for variants of the partial loss of flow accident. To treat uncertainties, three methods were used, ranging from conservative, deterministic methods, to more realistic and computationally demanding Monte Carlo-based methods. To facilitate the computational requirements of the thermal hydraulic analysis, a script-based interface was created for VIPRE. This scripted interface (written in Matlab) supplants the existing file-based interface. This interface allows for repeated, automatic execution of the VIPRE code on a script-modifiable input data, and parses and stores output data to disk. This endows the analyst with much greater power to use VIPRE in parametric studies, or using the Monte Carlo-based uncertainty analysis methodology. The Matlab environment also provides powerful visualization capability that greatly eases the task of data analysis. / by Stuart R. Blair. / Nucl.E.and S.M.

Thesis (Sc. D.)--Massachusetts Institute of Technology, Dept. of Nuclear Engineering, 2004. / Includes bibliographical references (p. 309-314). / A systematic, detailed major component and system design evaluation and multiple parameter optimization under practical constraints has been performed of the family of supercritical CO2 Brayton power cycles for application to advanced nuclear reactors. The recompression cycle is shown to excel with respect to simplicity, compactness, cost and thermal efficiency. The main advantage of the supercritical CO2 cycle is comparable efficiency with the helium Brayton cycle at significantly lower temperature (550⁰C vs. 850 ⁰C), but higher pressure (20 MPa vs. 8 MPa). The supercritical CO2 cycle is well suited to any type of nuclear reactor with core outlet temperature above [approx.] 500 ⁰C in either direct or indirect versions. By taking advantage of the abrupt property changes near the critical point of CO2 the compression work can be reduced, which results in a significant efficiency improvement. However, a real gas cycle requires much more careful optimization than an ideal gas Brayton cycle. Previous investigations by earlier authors were systematized and refined in the present work to survey several different CO2 cycle layouts. Inter- cooling, re-heating, re-compressing and pre-compressing were considered. The recompression cycle was found to yield the highest efficiency, while still retaining simplicity. Inter-cooling is not attractive for this type of cycle as it offers a very modest efficiency improvement. Re-heating has a better potential, but it is applicable only to indirect cycles. Economic analysis of the benefit of re-heating for the indirect cycle showed that using more than one stage of re-heat is economically unattractive. / (cont.) For the basic design, turbine inlet temperature was conservatively selected to be 550⁰C and the compressor outlet pressure set at 20 MPa. For these operating conditions the direct cycle achieves 45.3 % thermal efficiency and reduces the cost of the power plant by [approx.] 18% compared to a conventional Rankine steam cycle. The capital cost of the basic design compared to a helium Brayton cycle is about the same, but the supercritical CO2 cycle operates at significantly lower temperature. The current reactor operating experience with CO2 is up to 650⁰C, which is used as the turbine inlet temperature of an advanced design. The thermal efficiency of the advanced design is close to 50% and the reactor system with the direct supercritical CO2 cycle is - 24% less expensive than the steam indirect cycle and 7% less expensive than a helium direct Brayton cycle. It is expected in the future that high temperature materials will become available and a high performance design with turbine inlet temperatures of 700⁰C will be possible. This high performance design achieves a thermal efficiency approaching 53%, which yields additional cost savings. The turbomachinery is highly compact and achieves efficiencies of more than 90%. For the 600 MWth/246 MWe power plant the turbine body is 1.2 m in diameter and 0.55 m long, which translates into an extremely high power density of 395 MWe/m3. The compressors are even more compact as they operate close to the critical point where the density of the fluid is higher than in the turbine. The power conversion unit that houses these components and the generator is 18 m tall and 7.6 m in diameter. Its power density (MWe/m3) is about 46% higher than that of the helium GT-MHR ... / by Vaclav Dostal. / Sc.D.

Thesis (S.M.)--Massachusetts Institute of Technology, Dept. of Nuclear Engineering, 2004. / Includes bibliographical references (p. 75-77). / Fundamental objectives of sustainability, economics, safety and reliability, and proliferation resistance, physical protection and stakeholder relations must be considered during the design of an advanced reactor. However, an advanced reactor's core damage frequency dominates all other considerations at the preliminary stage of reactor design. An iterative four-step methodology to guide the MIT gas-cooled fast reactor emergency core cooling system design through PRA insights was utilized based upon the preliminary stage of the reactor design and activities currently ongoing in the nuclear industry, regulator, and universities regarding advanced reactors. Advanced reactor designs also face an uncertain regulatory environment. It was concluded from the move towards risk- informed regulations of current reactors, that there will be some level of probabilistic insights in the regulations and supporting regulatory documents for advanced, "Generation-IV" nuclear reactors. The four-step methodology is moreover used to help designers analyze designs under potential risk-informed regulations and predict design justifications the regulator will require during the licensing process. The iterative design guidance methodology led to a reduction of the CDF contribution due to a LOCA of over three orders of magnitude from the baseline ECCS design (from 1.19x10-5 to 6.48x10-8 for the 3x100% loop configuration) and potential ECCS licensing issues were identified. This illustrates the value of formal design guidance based upon PRA. / by Michael J. Delaney. / S.M.

Thesis (S.M.)--Massachusetts Institute of Technology, Dept. of Nuclear Engineering, 2004. / Includes bibliographical references (leaf 69). / An analysis was conducted to determine the behavior of a high pressure electric arc discharge inside the Plasmatron Fuel Reformer. The Plasmatron Fuel Reformer uses an electric arc to partially combust fuel in an internal combustion engine to increase efficiency and reduce emissions. Solutions of the conservation equations for the arc yield temperature, pressure and velocity profiles for arcs with 0.2-0.8 A currents. Acquired knowledge was used to predict arc radius and power delivered by the arc as a function of current. / by Barrett A. Burns. / S.M.

Thesis (Ph. D.)--Massachusetts Institute of Technology, Dept. of Nuclear Engineering, 2003. / Includes bibliographical references (p. 189-194). / A heterogeneous LWR core design, which employs a thorium/uranium once through fuel cycle, is optimized for good economics, wide safety margins, minimal waste burden and high proliferation resistance. The focus is on the Whole Assembly Seed and Blanket (WASB) concept, in which the individual seed and blanket regions each occupy one full-size PWR assembly in a checkerboard core configuration. A Westinghouse 4-loop 1150 MWe PWR was selected as the reference plant design. The optimized heterogeneous core, after several iterations, employs 84 seed assemblies and 109 blanket assemblies. Each assembly has the characteristic 17x17 rod array. The seed fuel is composed of 20 w/o enriched annular UO2 pellets. Erbium is used in the fresh seed to help regulate local power peaking and reduce soluble boron concentrations. Erbium was evenly distributed into all pin central holes except for the peripheral pins and four corner pins of each assembly where more erbium was used due to their higher power level. The blanket fuel is a mixture of 87% ThO2 - 13% UO2 by volume, where the uranium is enriched to 10 w/o. The blanket fuel pin diameter is larger than the seed fuel pin diameter. There are two separate fuel management flows: a standard three-batch scheme is adopted for the seed (18 month cycle length) and a single-batch for the blanket, which is to stay in the core for up to 9 seed cycles. The WASB core design was analyzed by well known tools in the nuclear industry. The neutronic analysis was performed using the Studsvik Core Management System (CMS), which consists of three codes: CASMO-4, TABLES-3 and SIMULATE-3. Thermal-hydraulic analysis was performed using EPRI's VIPRE-01. / (cont.) Fuel performance was analyzed using FRAPCON. The radioactivity and decay heat from the spent seed and blanket fuel were studied using MIT's MCODE (which couples MCNP and ORIGEN) to do depletion calculations, and ORIGEN to analyze the spent fuel characteristics after discharge. The analyses show that the WASB core can satisfy the requirements of fuel cycle length and safety margins of conventional PWRs. The coefficients of reactivity are comparable to currently operating PWRs. However, the reduction in effective delayed neutron fraction (eff) requires careful review of the control systems because of its importance to short term power transients. Whole core analyses show that the total control rod worth of the WASB core is about 1/3 less than those of a typical PWR for a standard arrangement of Ag-In-Cd control rods in the core. The use of enriched boron in the control rods can effectively improve the control rod worth. The control rods have higher worth in the seed than in the blanket. Therefore, a new loading pattern has been designed so that almost all the control rods will be located in seed assemblies. However, the new pattern requires a redesign of the vessel head of the reactor, which is an added cost in case of retrofitting in existing PWRs. Though the WASB core has high power peaking factors, acceptable MDNBR in the core can be achieved under conservative assumptions by using grids with large local pressure loss coefficient in the blanket. However, the core pressure drop will increase by 70% ... / by Dean Wang. / Ph.D.

Thesis (Ph. D.)--Massachusetts Institute of Technology, Dept. of Nuclear Engineering, 1986. / MICROFICHE COPY AVAILABLE IN ARCHIVES AND SCIENCE / Bibliography: leaves 120-123. / by Anthony V. D'Amico. / Ph.D.

Thesis (S.M.)--Massachusetts Institute of Technology, Dept. of Nuclear Engineering, 2004. / Includes bibliographical references (leaves 152-155). / The international maritime container trade, which imports an average of 19,000 largely uninspected cargo containers to United States ports each day, has been identified as a potential avenue of attack for nuclear terrorism. Currently envisioned and deployed defensive measures that seek to detect and interdict concealed fissile material once containers have already reached a U.S. port do not adequately protect against nuclear threats due to the unique power and range of nuclear weapon effects. This thesis describes and examines a novel "ship-based" approach to container-borne fissile material detection where suites of radiation detectors with imaging capabilities are enclosed in standard, non-descript cargo containers and shipped in limited numbers aboard commercial containerships. Outfitted with communication hardware, these dedicated containerized units could provide crucial advance detection and notification of an inbound nuclear threat while the danger is still safely removed from U.S. shores. Attributes of the container shipping trade that would impact the performance and viability of the proposed ship-based approach were identified and investigated. / (cont.) Average available count times, based on the duration of shipping voyages, for container imports to representative ports on the east and west coasts of the U.S. where found to be 19.2 days and 13.3. days, respectively. These long count times will enhance the ability of the ship- based approach to confidently detect heavily shielded and well-concealed fissile material. A distribution for the average distributed density of commercial cargo, which affects radiation attenuation between the source and detectors, was also derived and found to have a favorably low mean value of 0.198 g/cm³. The coverage efficiency (i.e. the number of containerized units required to provide detection coverage over a given percentage of a reference vessel) variations associated with prospective modes of deployment were also investigated using Matlab- based computer simulations. Evaluated deployment strategies ranged from fully random placement of detection units to completely constrained optimal placement. Despite holding important advantages in terms of stealth, random deployment was found to require an average of between 2.2 to 3.3 times more detectors than optimal deployment, depending on the desired level of detection coverage. / (cont.) This result suggests that some combination of random and constrained deployment might yield an optimized balance between stealth and coverage efficiency. This analysis also identified significant efficiency and deployment flexibility benefits associated with units that could detect sources at ranges equal to, or greater than, 70 ft (21.3 m). Overall, no results were obtained that seriously challenged the potential efficacy and viability of the proposed ship-based approach. / by Brett P. Broderick. / S.M.

Thesis (Ph. D.)--Massachusetts Institute of Technology, Dept. of Nuclear Engineering, 2004. / Includes bibliographical references (leaves 152-154). / This work was aimed at reproducing experimental results in impurity compression of Ar, as well as the screening of recycling and non-recycling impurities from reaching the core plasma. As part of the study the code was upgraded in order to track the impurity flow from source till it reaches the core, include the energy dependence on recycled impurity atoms, and allow for more realistic impurity recycling in the SOL. This added capability allows the determination of which source locations are dominant in determining the core impurity level, where they cross the separatrix into the core and where they leave the core. The modeling reproduces within a factor of 2 the experimentally observed compression of Ar in the divertor of Alcator C-Mod. In addition it was found that under attached conditions recycling at the outer plasma edge (limiters located there) was the dominant source of Ar ions reaching the core (over 60%). For detached conditions divertor recycling replaces the outer edge in supplying the majority of Ar ions reaching the core. / (cont.) There appear to be two general flow patterns of impurities through the core plasma: Outboard launched impurities enter the core at the outside edge and flow out of the core on the inboard edge; Divertor launched impurities enter the core just outboard of the x-point and return to the divertor just inboard of the x-point. The study of non-recycling impurities was also carried out and it was found that the penetration factor (PF) for outboard-launched impurities (Carbon was used as the prototype) were a factor of 3 times more likely to reach the core than inboard-launched impurities (experimental result gave the ratio as 20). Increasing the background SOL plasma flow to the experimental levels doubles the model ratio and other factors capable of reducing the discrepancy are studied. Thus the experimental poloidal variation in PF is qualitatively reproduced. Values of PF for recycling impurities (a global quantity) matched the experimental magnitudes when experimental values for SOL flow were used. / by Taekyun Chung. / Ph.D.

Thesis (S.M.)--Massachusetts Institute of Technology, Dept. of Nuclear Engineering, 2002. / Includes bibliographical references (p. 92-94). / The construction of a next generation fast nuclear reactor using liquid lead-bismuth as a coolant demands the design of applied technology to remove 210Po from the neutron activated lead-bismuth eutectic. Experiments were performed to determine the kinetics of polonium hydride and lead-polonide released from molten lead-bismuth to determine the rate response of gaseous polonium chemical species in contact with various argon and hydrogen gas streams. It was determined that the rate release of polonium hydride is slightly higher at lower temperatures. The kinetic response is also faster with increased hydrogen content, as evident by the determined equilibrium constant. In addition, experiments involving the adsorption of gaseous polonium species on metallic praseodymium were undertaken. Formation of an oxidation layer and physical deterioration of the praseodymium inhibited proper sample analysis. The extraction techniques of rare-earth filtering and polonium hydride stripping and their relative figures of merit were discussed. Of the two techniques, a small-scale design adopting polonium hydride stripping was explored to address basic issues of design, fabrication, operation, and maintenance of an online polonium extraction system. Pending results of further investigation on alkaline extraction and electro-deposition experiments a small-scale design may be pursued. / by Christopher L. Larson. / S.M.