In the process of exploring the thorium fuel application in CANDU reactors, it is important to consider the proliferation resistance level as a parameter for comparison with current natural uranium fuel. The concept of a whole fuel cycle was introduced to show the variations in the proliferation resistance level as the material is flowing through the cycle. The depletion and decay histories were simulated with SCALE 6.1 code and the results such as isotopes composition, decay heat, and radioactivity were used to analyze the material attractiveness of pure heavy metal for weapon production. They also served as the intrinsic features during the proliferation resistance level calculation. The Multi-Attribute Utility Analysis (MAUA) method developed by Chalton was used to compare different CANDU fuel cycles with quantified values (PR) from the viewpoint of proliferation resistance. To improve the biased MAUA results that gave a PR of 0.76 to CANDU while 0.93 to PWR, the attributes of size/weight and refueling scheme were reconsidered. In addition, the sensitive technology involved was added for the proliferation resistance recalculation. The results showed an increased PR value of 0.82 for natural uranium CANDU reactor as well as a decreasing trend of PR at the back end. PWR has a PR of 0.82 with revised MAUA method. The PR comparison of thorium and natural uranium fuel indicated that Th/Pu fuel has a slightly higher PR value in the reactor. The Figure of Merit (FOM) method developed by Bathke was used to validate the PR results from another perspective: the attractiveness of pure heavy metals that are suitable for nuclear weapon production. The results showed that FOM of plutonium keeps increasing with decay time and the trend becomes more significant after disposal in the deep geological repository. The FOM of uranium from Th/Pu cases is higher than that of Pu within several hundred years but maintains a decreasing trend. The decreasing FOM of uranium is preferred for direct disposal in deep geological repository. The decreased PR level and the increased FOM value of plutonium at the back end of a fuel cycle indicate the importance of implementing the security and safeguard for each facility dealing with nuclear materials. The comparison results of PR and FOM values for different fuel provided feedback and suggestions for the new fuel application. / Thesis / Master of Applied Science (MASc)

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Made available in DSpace on 2014-10-09T12:52:33Z (GMT). No. of bitstreams: 0 / Made available in DSpace on 2014-10-09T14:02:23Z (GMT). No. of bitstreams: 0 / Os acidentes causados pelo efeito dominó são dos mais graves ocorridos na indústria química e de processo. Mesmo sendo o potencial destrutivo desses eventos acidentais bastante conhecido, pouca atenção tem sido dada a este problema pela literatura técnica e uma metodologia completa e aprovada para a avaliação quantitativa da contribuição do efeito dominó ao risco industrial ainda não está plenamente desenvolvida. O presente estudo propõe um procedimento sistemático para a avaliação quantitativa do efeito dominó em plantas químicas do ciclo do combustível nuclear. O trabalho é baseado em avanços recentes feitos na modelagem de danos a equipamentos de processo causados por incêndios e explosões devido aos vetores de propagação (radiação de calor, sobrepressão e projeção de fragmentos). Dados disponíveis na literatura técnica e novos modelos de vulnerabilidade deduzidos para diversas categorias de equipamentos de processo foram utilizados no presente trabalho. O procedimento proposto é aplicado a uma área de tancagem típica de uma planta de reconversão situada em um sítio que abriga varias outras instalações do ciclo do combustível nuclear. São analisados os vários eventos iniciadores, seus vetores de propagação, as conseqüências desses eventos e as freqüências associadas ao efeito dominó. / Dissertação (Mestrado) / IPEN/D / Instituto de Pesquisas Energeticas e Nucleares - IPEN/CNEN-SP

The fluoride-salt-cooled high-temperature reactor (FHR) is a novel reactor design benefitting from passive safety features, high operating temperatures with corresponding high conversion efficiency, to name a few key features. The fuel is a layered graphite plank configuration containing enriched uranium oxycarbide (UCO) tri-structural isotropic (TRISO) fuel particles. Fuel cycle cost (FCC) models have been used to analyze and optimize fuel plate thicknesses, enrichment, and packing fraction as well as to gauge the economic competitiveness of this reactor design. Since the development of the initial FCC model, many corrections and modifications have been identified that will make the model more accurate. These modifications relate to corrections made to the neutronic simulations and the need for a more accurate fabrication costs estimate. The former pertains to a MC Dancoff factor that corrects for fuel particle neutron shadowing that occurs for double-heterogeneous fuels in multi-group calculations. The latter involves a detailed look at the fuel fabrication process to properly account for material, manufacturing, and quality assurance cost components and how they relate to the heavy metal loading in a FHR fuel plank. It was found that the fabrication cost may be a more significant portion of the total FCC than was initially attributed. TRISO manufacturing cost and heavy metal loading via packing fraction were key factors in total fabrication cost. This study evaluated how much neutronic and fabrication cost corrections can change the FCC model, optimum fuel element parameters, and the economic feasibility of the reactor design.

Uncertainty surrounds the future evolution of key factors affecting the attractiveness of various nuclear fuel cycles, rendering the concept of a unique optimal fuel cycle transition strategy invalid. This work applies decision-making under uncertainty to fuel cycle transition analysis, demonstrating a new, systematic methodology for choosing flexible, adaptable hedging strategies that yield middle-of-the-road results until uncertainties are resolved. A case study involving transition from the current once-through light water reactor (LWR) fuel cycle to one relying on continuous recycle in fast reactors (FRs) is cast as a no-data decision problem. The transition is subject to uncertainty in the cost of spent nuclear fuel (SNF) and high-level waste (HLW) disposal in a geologic repository, slated to open some years into the future. Following the repository open date, the cost of SNF and HLW disposal is made known, and may take on one of five possible values. Strategies for the transition are enumerated and simulated using VEGAS, a systems model of the nuclear fuel cycle that solves for its material balance and applies input cost data to calculate the associated annual levelized cost of electricity (LCOE). Perfect information strategies are found using the lowest average, maximum, and integrated LCOE objective functions. The loss in savings for following a strategy other than the perfect information strategy is the “regret” which is calculated by evaluating the performance of each strategy for every end-state. Hedging strategies are then selected by either minimizing the maximum or the expected regret. Generally, the optimal hedging strategy identified using the decision methodology suggests a partial transition to a closed fuel cycle prior to the repository open date. Once the repository opens, the transition may be abandoned or accelerated depending on which disposal cost outcome is realized. The lowest average and integrated LCOE objective functions perform similarly; however, the lowest maximum LCOE objective function appears overly sensitive to aberrations in the annual LCOE that arise due to idle reprocessing capacity. The minimax regret choice criterion is shown to be more conservative than the lowest expected regret choice criterion, as it acts to hedge against the worst-case outcome. By following a hedging strategy, agents may alter their fuel cycle strategy more readily once uncertainties are resolved. This results since hedging strategies provide flexibility in the nuclear fuel cycle, preserving what options exist. To this end, the work presented here may provide guidance for agent-based, behavioral modeling in fuel cycle simulators, as well as decision-making in real world applications. / text

The sintering behavior of uranium and uranium-zirconium alloys in the alpha phase were characterized in this research. Metal uranium powder was produced from pieces of depleted uranium metal acquired from the Y-12 plant via hydriding/dehydriding process. The size distribution and morphology of the uranium powder produced by this method were determined by digital optical microscopy. Once the characteristics of the source uranium powder were known, uranium and uranium-zirconium pellets were pressed using a dual-action punch and die. The majority of these pellets were sintered isothermally, first in the alpha phase near 650°C, then in the gamma phase near 800°C. In addition, a few pellets were sintered using more exotic temperature profiles. Pellet shrinkage was continuously measured in situ during sintering. The isothermal shrinkage rates and sintering temperatures for each pellet were fit to a simple model for the initial phase of sintering of spherical powders. The material specific constants required by this model, including the activation energy of the process, were determined for both uranium and uranium-zirconium. Following sintering, pellets were sectioned, mounted, and polished for imaging by electron microscopy. Based on these results, the porosity and microstructure of the sintered pellets were analyzed. The porosity of the uranium-zirconium pellets was consistently lower than that of the pure uranium pellets. In addition, some formation of an alloyed phase of uranium and zirconium was observed. The research presented within this thesis is a continuation of a previous project; however, this research has produced many new results not previously seen. In addition, a number of issues left unresolved by the previous project have been addressed and solved. Most notably, the low original output of the hydride/dehydride powder production system has been increased by an order of magnitude, the actual characteristics of the powder have been measured and determined, shrinkage data was successfully converted into a sintering model, an alloyed phase of uranium and zirconium was produced, and pellet cracking due to delamination has been eliminated.

Nuclear fuel cycles (NFC) are the collection of interconnected processes which generate electricity through nuclear power. Due to the high degree of coupling between components even in the simplest cycles, the need for a dynamic fuel cycle simulator and analysis framework arises. The work presented herein develops essential physics models of nuclear power reactors and incorporate them into a NFC simulation framework. First, a one-energy group reactor model is demonstrated. This essential physics model is then to simulate a sampling fuel cycles which are perturbations of well known base-case cycles. Because the NFC may now be simulated quickly, stochastically modeling many fuel cycle realizations dramatically expands the parameter space which may be analyzed. Finally, a multigroup reactor model which incorporates spectral changes as a function of burnup is presented to increase the fidelity of the original one-group reactor. These methods form a suite of modeling technologies which reach from the lowest levels (individual components) to the highest (inter-cycle comparisons). Prior to the development of this model suite, such broad-ranging analysis had been unrealistic to perform. The work here thus presents a new, multi-scale approach to fuel cycle system design. / text

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Made available in DSpace on 2014-10-09T12:35:33Z (GMT). No. of bitstreams: 0 / Made available in DSpace on 2014-10-09T14:04:01Z (GMT). No. of bitstreams: 0 / Tese (Doutoramento) / IPEN/T / Instituto de Pesquisas Energeticas e Nucleares - IPEN-CNEN/SP

Made available in DSpace on 2014-10-09T12:55:29Z (GMT). No. of bitstreams: 0 / Made available in DSpace on 2014-10-09T14:04:54Z (GMT). No. of bitstreams: 0 / Tese (Doutoramento) / IPEN/T / Instituto de Pesquisas Energeticas e Nucleares - IPEN-CNEN/SP