Gas hydrate investigations are converging on a suite of common techniques for hydrate observation and quantification. Samples retrieved and analyzed at full in situ pressures are the ”gold standard” with which the physical and chemical analysis of conventional cores, as well as the interpretation of geophysical data, are calibrated and groundtruthed. Methane mass balance calculations from depressurization of pressure cores provide the benchmark for gas hydrate concentration assessment. Nondestructive measurements of pressure cores have removed errors in the estimation of pore volume, making this methane mass balance technique accurate and robust. Data from methane mass balance used to confirm chlorinity baselines makes porewater freshening analysis more accurate. High-resolution nondestructive analysis of gas-hydratebearing cores at in situ pressures and temperatures also provides detailed information on the in situ nature and morphology of gas hydrate in sediments, allowing better interpretation of conventional core thermal images as well as downhole electrical resistivity logs. The detailed profiles of density and Vp, together with spot measurements of Vs, electrical resistivity, and hardness, provide background data essential for modeling the behavior of the formation on a larger scale. X-ray images show the detailed hydrate morphology, which provides clues to the mechanism of deposit formation and data for modeling the kinetics of deposit dissociation. Gashydrate- bearing pressure cores subjected to X-ray tomographic reconstruction provide evidence that gas hydrate morphology in many natural sedimentary environments is particularly complex and impossible to replicate in the laboratory. Even when only a small percentage of the sediment column is sampled with pressure cores, these detailed measurements greatly enhance the understanding and interpretation of the more continuous data sets collected by conventional coring and downhole logging. Pressure core analysis has become the keystone that links these data sets together and is an essential component of modern gas hydrate investigations.

This study presents a stratigraphic investigation of the marine sediment core MD04-2822 from the Rockall Trough (56° 50.54' N, 11° 22.96' W; 2344 m water depth). This core is currently the only available high resolution record for the calibration of Late Quaternary sedimentary sequences of the British (Hebridean) margin. It therefore offers an unprecedented archive of changing sedimentological and climatological conditions for the last 175,000 years. The high resolution, multi-proxy records have enabled surface and deep water conditions within the Rockall Trough to be reconstructed. In addition, the fluctuating nature of ice-rafted debris (IRD) inputs to the MD04-2822 site allows a first order attempt of BIS dynamics for the entirety of the last glacial period (i.e. from the demise of the last interglacial to the decay of the Devensian/Weichselian ice sheet) as well as the majority of the penultimate (Saalian/MIS 6) glaciation. Sediment core MD04-2822 is ideally located to capture the dynamics of the British Ice Sheet (BIS) via a continuous record of IRD and fine-grained terrigenous inputs. Fundamental to this is the construction of a robust chronology. This was achieved via: the correlation of the benthic δ¹⁸O record to a global δ¹⁸O stack (SPECMAP); the correlation of the surface proxies (% N. pachyderma (sinistral) and XRF Ca) to the Greenland δ¹⁸O and Antarctic methane ice core records; and radiocarbon dating. This chronology was validated using both radiocarbon dating and tephra horizons. An evaluation of the event stratigraphy approach used in the construction of the MD04-2822 chronology is presented. The marine record provides a valuable archive of past ice sheet dynamics as much terrestrial evidence is removed or obscured by subsequent ice sheet oscillations MD04-2822 provides the first evidence for the expansion of the BIS onto the Hebridean Margin during MIS6 (thereby confirming previous long-range seismic correlations). The continuous sedimentation at MD04-2822 enabled the first insights into the early dynamics of the last BIS. Increases in IRD and fine grained terrigenous material delivered to the MD04-2822 at ca. 72 kyr represent the first significant delivery of material from the BIS across the continental shelf to the core site. The BIS would therefore have attained a marine calving margin by this time. A multi-proxy investigation of provenance was undertaken, however unequivocal provenance determinations remain problematic. The location of the core suggest the proximal BIS as the most likely source of terrigenous inputs. The expanded nature of the MD04-2822 sediments during the penultimate deglacial (Termination II) provides the first details of BIS dynamics for this period: the interplay of large inputs of freshwater from the decay of the Saalian (MIS 6) ice sheets (including the BIS) upon the surface and deep water circulation of the North Atlantic is investigated. In addition, sub-orbital climatic variability is documented at this location throughout the last interglacial (MIS 5e) and appears to be an intrinsic feature of both the N.E. Atlantic surface and deep water circulation of the last 175 kyr.

The recent and rapid growth of horizontal drilling in the Anadarko basin necessitates newer studies to characterize reservoir and source rock quality in the region. Most oil production in the basin comes from the Granite Wash reservoirs, which are composed of stacked tight sandstones and conglomerates that range from Virgillian (305-299 Ma) to Atokan (311-309.4 Ma) in age. By utilizing geophysical well logging data available in raster format, the Granite Wash reservoirs and their respective marine flooding surfaces were stratigraphically mapped across the regional fault systems. Additionally, well log trends were calibrated with coincident core data to minimize uncertainty regarding facies variability and lateral continuity of these intervals. In this thesis, inferred lithofacies were grouped into medium submarine fan lobe, distal fan lobe, and offshore facies (the interpreted depositional environments). By creating isopach and net sand maps in Petra, faulting in the Missourian was determined to have occurred syndepositionally at the fifth order scale of stratigraphic hierarchy.

Advanced computational tools are applied to simulate a nuclear power plant (NPP) control rod assembly ejection (CRE) accident. The impact of these reactivity-initiated accidents (RIAs) on core reactivity behavior, 3D power distribution and stochastic reactivity estimation are evaluated. The three tools used are: the thermal-hydraulic (TH) RELAP5 (R5) code, the neutronic (NK) PARCS (P3D) code, and the coupled version P3D/R5, with specially developed linkage using the environment code MATLAB. This study considers three different-size cores: NPP1 (2772 MWt); NPP2 (530 MWt); and NPP3 (1061 MWt). The three cores have the same general design and control rod assembly (CRA) positions, and the ejected CRA has similar worth and at the same rod ejection pace. The CRE is assessed under both hot zero power (HZP) and hot full power (HFP) conditions. The analyses indicate that RIA modeling and simulation should be carried out through a systematic coding and configuration approaches, otherwise the results will not capture the true transient behavior of the core under analysis. The simulation of one code depends on the appropriate configuration of parameters generated by the other code and on the correct determination of the TH/NK mapping weight factors for the various mesh regions in each of the models. From the design point of view, the standalone codes predict milder magnitude of power and reactivity increase compared to the coupled P3D/R5 simulation. The magnitudes of reduced peak power and reactivity become larger as the core size shrinks. The HFP simulation shows that the three NPPs have the same transient peak value, but the post-transient steady power is lower for a smaller core. The HZP analysis indicates that the transient peak is lower for the smaller core, but the post-transient power occurs at the same level. The three-dimensional (3D) power distributions are different among the HFP and HZP cases, but do not depend on the size of the core. The results indicate: i) HFP: core power increases in the area surrounding the ejected rod/bank assembly, and this increase becomes lower as the NPPs shrinks however, the power is well-distributed after the transient; and ii) HZP: the area surrounding the CRA stays hotter, but the 3D peak assembly factor becomes lower, during and after the transients, as the NPPs shrinks. These features confirm that the smaller cores yield a safer response to a given inserted reactivity compared to larger cores. A stochastic extended Kalman filter (EKF) algorithm is implemented to estimate the reactivity based on the reactor power profile, after the addition of random noise. The inverse point kinetics (IPK) deterministic method is also implemented and the results of the application of EKF and IPK are compared to the P3D/R5 simulation. The following sophisticated strategies made the EKF algorithm robust and accurate: the system is modeled by a set of continuous time nonlinear stochastic differential equations; the code uses a time step directly based on the power measured and applies that to the model for online discretization and linearization; filter tuning goes automatically up from the first time step; and the state noise covariance matrix is updated online at each time step. It was found that the IPK reactivity has higher noise content compared to the EKF reactivity for all cases. Thus, the EKF presents superior and more accurate results. Furthermore, under a small reactivity insertion, the IPK reactivity varies widely from positive to negative values: this variation is not observed within the EKF. A sensitivity analysis for three distinct standard deviation (SD) noise measurements suggests that EKF is superior to IPK method, independent of the noise load magnitude. As the noise content increases, the error between the IPK and P3D/R5 reactivity also increases. A sensitivity analysis for five distinct carry-over effects of different random noise loads indicates that the random addition of different noise loads to the reactor power does not change the overall performance of both algorithms. / Este trabalho aplica métodos computacionais avançados para simular a ejeção de barras de controle (CRE) em uma planta térmica nuclear (NPP). São avaliados o impacto da ocorrência de acidentes iniciados por reatividade (RIAs) na reatividade total, na distribuição da potência em três dimensões (3D) e na determinação da reatividade. As ferramentas utilizadas são: o código termo-hidráulico (TH) RELAP5 (R5), o código neutrônico (NK) PARCS (P3D), a versão acoplada P3D/R5, e o ambiente computacional MATLAB. Este estudo considera três reatores nucleares de diferentes tamanhos: NPP1 (2772 MWT); NPP2 (530 MWt); e NPP3 (1061 MWt). Os três núcleos possuem projeto similar e idêntica posição dos grupos das barras de controle (CRA), além do mesmo valor de reatividade diferencial das CRA ejetadas e idêntica velocidade de ejeção. A ocorrência da CRE é avaliada sob condições de hot zero power (HZP) e de hot full power (HFP). As análises indicam que a modelagem e a simulação de RIAs devem ser realizadas sistematicamente, caso contrário os resultados não irão refletir o comportamento em regime transitório do núcleo. A simulação de um modelo em um código depende da apropriada configuração de parâmetros gerados pelo outro código e da determinação adequada do mapeamento TH/NK para as várias malhas dos modelos. Do ponto de vista de projeto, a utilização de códigos independentes resulta em cálculos de potência e reatividade conservadores em comparação com os resultados utilizando-se P3D/R5. Os picos de potência e de reatividade são menores à medida que o núcleo encolhe. A simulação em condições de HFP resulta em valores de pico de potência similares durante transitório para as três NPPs, mas a potência de pós-transitórios é menor para o menor núcleo. A análise em condições de HZP também indica que o valor máximo durante o transitório é menor para o menor núcleo, mas o pós-transitórios ocorre aos mesmos níveis de potência das demais NPPS. A distribuição de potência em 3D também apresenta resultados distintos para condições de HFP e HZP, mas tais resultados são independentes do tamanho do núcleo: i) HFP: há um aumento da potência do núcleo em torno da CRE, mas tal comportamento diminui para núcleos menores - no entanto, a potência é bem distribuída após o transitório; e ii) HZP: há aumento de potência na área do CRE, mas o pico de potência em 3D é menor durante e depois dos transitórios para núcleos menores. Tais características indicam que os núcleos menores respondem de forma mais segura quando da inserção de reatividade em comparação a reatores de maiores dimensões. O método estocástico de filtragem de Kalman estendido (EKF) foi codificado para estimar a reatividade com base no perfil de potência da NPP, após a adição de ruído aleatório. O método determinístico da cinética pontual inversa (IPK) também foi implementado e os resultados da aplicação dos algoritmos do EKF e IPK foram comparados com os resultados da simulação do P3D/R5. As seguintes estratégias, implementadas neste trabalho, possibilitaram a aplicação robusta e precisa do EKF: o sistema foi modelado por um conjunto de equações diferenciais não-lineares estocásticas de tempo contínuo; o algoritmo obtém o passo de tempo diretamente da potência medida e aplica-o ao modelo para a discretização e linearização online; o ajuste do filtro ocorre automaticamente a partir do primeiro passo de tempo; e a matriz de covariância do ruído no estado é atualizada online. Verificou-se que a reatividade calculada pelo método IPK possui maior nível de ruído quando comparada ao EKF para todos os casos estudados. Portanto, o EKF apresenta resultados superiores e mais precisos. Além disso, sob uma pequena inserção de reatividade, a reatividade calculada pelo método IPK varia consideravelmente de valores positivos para negativos: esta variação não é observada com o EKF. Uma análise de sensibilidade para três desvios padrão (SD) sugere que o algoritmo EKF é superior ao método IPK, independente da magnitude do ruído. Com o aumento da magnitude do ruído, o erro entre as reatividades calculadas pelo IPK e pelo P3D/R5 aumenta. A análise de sensibilidade para cinco ruídos aleatórios indica que a adição de ruído na potência do reator não altera o desempenho global de ambos os algoritmos.

Advanced computational tools are applied to simulate a nuclear power plant (NPP) control rod assembly ejection (CRE) accident. The impact of these reactivity-initiated accidents (RIAs) on core reactivity behavior, 3D power distribution and stochastic reactivity estimation are evaluated. The three tools used are: the thermal-hydraulic (TH) RELAP5 (R5) code, the neutronic (NK) PARCS (P3D) code, and the coupled version P3D/R5, with specially developed linkage using the environment code MATLAB. This study considers three different-size cores: NPP1 (2772 MWt); NPP2 (530 MWt); and NPP3 (1061 MWt). The three cores have the same general design and control rod assembly (CRA) positions, and the ejected CRA has similar worth and at the same rod ejection pace. The CRE is assessed under both hot zero power (HZP) and hot full power (HFP) conditions. The analyses indicate that RIA modeling and simulation should be carried out through a systematic coding and configuration approaches, otherwise the results will not capture the true transient behavior of the core under analysis. The simulation of one code depends on the appropriate configuration of parameters generated by the other code and on the correct determination of the TH/NK mapping weight factors for the various mesh regions in each of the models. From the design point of view, the standalone codes predict milder magnitude of power and reactivity increase compared to the coupled P3D/R5 simulation. The magnitudes of reduced peak power and reactivity become larger as the core size shrinks. The HFP simulation shows that the three NPPs have the same transient peak value, but the post-transient steady power is lower for a smaller core. The HZP analysis indicates that the transient peak is lower for the smaller core, but the post-transient power occurs at the same level. The three-dimensional (3D) power distributions are different among the HFP and HZP cases, but do not depend on the size of the core. The results indicate: i) HFP: core power increases in the area surrounding the ejected rod/bank assembly, and this increase becomes lower as the NPPs shrinks however, the power is well-distributed after the transient; and ii) HZP: the area surrounding the CRA stays hotter, but the 3D peak assembly factor becomes lower, during and after the transients, as the NPPs shrinks. These features confirm that the smaller cores yield a safer response to a given inserted reactivity compared to larger cores. A stochastic extended Kalman filter (EKF) algorithm is implemented to estimate the reactivity based on the reactor power profile, after the addition of random noise. The inverse point kinetics (IPK) deterministic method is also implemented and the results of the application of EKF and IPK are compared to the P3D/R5 simulation. The following sophisticated strategies made the EKF algorithm robust and accurate: the system is modeled by a set of continuous time nonlinear stochastic differential equations; the code uses a time step directly based on the power measured and applies that to the model for online discretization and linearization; filter tuning goes automatically up from the first time step; and the state noise covariance matrix is updated online at each time step. It was found that the IPK reactivity has higher noise content compared to the EKF reactivity for all cases. Thus, the EKF presents superior and more accurate results. Furthermore, under a small reactivity insertion, the IPK reactivity varies widely from positive to negative values: this variation is not observed within the EKF. A sensitivity analysis for three distinct standard deviation (SD) noise measurements suggests that EKF is superior to IPK method, independent of the noise load magnitude. As the noise content increases, the error between the IPK and P3D/R5 reactivity also increases. A sensitivity analysis for five distinct carry-over effects of different random noise loads indicates that the random addition of different noise loads to the reactor power does not change the overall performance of both algorithms. / Este trabalho aplica métodos computacionais avançados para simular a ejeção de barras de controle (CRE) em uma planta térmica nuclear (NPP). São avaliados o impacto da ocorrência de acidentes iniciados por reatividade (RIAs) na reatividade total, na distribuição da potência em três dimensões (3D) e na determinação da reatividade. As ferramentas utilizadas são: o código termo-hidráulico (TH) RELAP5 (R5), o código neutrônico (NK) PARCS (P3D), a versão acoplada P3D/R5, e o ambiente computacional MATLAB. Este estudo considera três reatores nucleares de diferentes tamanhos: NPP1 (2772 MWT); NPP2 (530 MWt); e NPP3 (1061 MWt). Os três núcleos possuem projeto similar e idêntica posição dos grupos das barras de controle (CRA), além do mesmo valor de reatividade diferencial das CRA ejetadas e idêntica velocidade de ejeção. A ocorrência da CRE é avaliada sob condições de hot zero power (HZP) e de hot full power (HFP). As análises indicam que a modelagem e a simulação de RIAs devem ser realizadas sistematicamente, caso contrário os resultados não irão refletir o comportamento em regime transitório do núcleo. A simulação de um modelo em um código depende da apropriada configuração de parâmetros gerados pelo outro código e da determinação adequada do mapeamento TH/NK para as várias malhas dos modelos. Do ponto de vista de projeto, a utilização de códigos independentes resulta em cálculos de potência e reatividade conservadores em comparação com os resultados utilizando-se P3D/R5. Os picos de potência e de reatividade são menores à medida que o núcleo encolhe. A simulação em condições de HFP resulta em valores de pico de potência similares durante transitório para as três NPPs, mas a potência de pós-transitórios é menor para o menor núcleo. A análise em condições de HZP também indica que o valor máximo durante o transitório é menor para o menor núcleo, mas o pós-transitórios ocorre aos mesmos níveis de potência das demais NPPS. A distribuição de potência em 3D também apresenta resultados distintos para condições de HFP e HZP, mas tais resultados são independentes do tamanho do núcleo: i) HFP: há um aumento da potência do núcleo em torno da CRE, mas tal comportamento diminui para núcleos menores - no entanto, a potência é bem distribuída após o transitório; e ii) HZP: há aumento de potência na área do CRE, mas o pico de potência em 3D é menor durante e depois dos transitórios para núcleos menores. Tais características indicam que os núcleos menores respondem de forma mais segura quando da inserção de reatividade em comparação a reatores de maiores dimensões. O método estocástico de filtragem de Kalman estendido (EKF) foi codificado para estimar a reatividade com base no perfil de potência da NPP, após a adição de ruído aleatório. O método determinístico da cinética pontual inversa (IPK) também foi implementado e os resultados da aplicação dos algoritmos do EKF e IPK foram comparados com os resultados da simulação do P3D/R5. As seguintes estratégias, implementadas neste trabalho, possibilitaram a aplicação robusta e precisa do EKF: o sistema foi modelado por um conjunto de equações diferenciais não-lineares estocásticas de tempo contínuo; o algoritmo obtém o passo de tempo diretamente da potência medida e aplica-o ao modelo para a discretização e linearização online; o ajuste do filtro ocorre automaticamente a partir do primeiro passo de tempo; e a matriz de covariância do ruído no estado é atualizada online. Verificou-se que a reatividade calculada pelo método IPK possui maior nível de ruído quando comparada ao EKF para todos os casos estudados. Portanto, o EKF apresenta resultados superiores e mais precisos. Além disso, sob uma pequena inserção de reatividade, a reatividade calculada pelo método IPK varia consideravelmente de valores positivos para negativos: esta variação não é observada com o EKF. Uma análise de sensibilidade para três desvios padrão (SD) sugere que o algoritmo EKF é superior ao método IPK, independente da magnitude do ruído. Com o aumento da magnitude do ruído, o erro entre as reatividades calculadas pelo IPK e pelo P3D/R5 aumenta. A análise de sensibilidade para cinco ruídos aleatórios indica que a adição de ruído na potência do reator não altera o desempenho global de ambos os algoritmos.

No description available.

Perhaps surprisingly, the largest experience in operating nuclear power plants has been in nuclear naval propulsion, particularly submarines. This accumulated experience may become the basis of a proposed new generation of compact nuclear power plant designs. In an effort to de-carbonise commercial freight shipping, there is growing interest in the possibility of using nuclear propulsion systems. Reactor cores for such an application would need to be fundamentally different from land-based power generation systems, which require regular refueling, and from reactors used in military submarines, as the fuel used could not conceivably be as highly enriched. Nuclear-powered propulsion would allow ships to operate with low fuel costs, long refueling intervals, and minimal emissions; however, currently such systems remain largely confined to military vessels. This research project undertakes computational modeling of possible soluble-boron-free (SBF) reactor core designs for this application, with a view to informing design decisions in terms of choices of fuel composition, materials, core geometry and layout. Computational modeling using appropriate reactor physics (e.g. WIMS, MONK, Serpent and PANTHER), thermal-hydraulics etc. codes (e.g. COBRA-EN) is used for this project. With an emphasis on reactor physics, this study investigates possible fuel assembly and core designs for civil marine propulsion applications. In particular, it explores the feasibility of using uranium/thorium-rich fuel in a compact, long-life reactor and seek optimal choices and designs of the fuel composition, reactivity control, assembly geometry, and core loading in order to meet the operational needs of a marine propulsion reactor. In this reactor physics and 3D coupled neutronics/thermal-hydraulics study, we attempt to design a civil marine reactor core that fulfills the objective of providing at least 15 effective full-power-years (EFPY) life at 333 MWth. In order to unleash the benefit of thorium in a long life core, the micro-heterogeneous ThO2-UO2 duplex fuel is well-positioned to be utilized in our proposed civil marine core. Unfortunately, A limited number of studies of duplex fuel are available in the public domain, but its use has never been examined in the context of a SBF environment for long-life small modular rector (SMR) core. Therefore, we assumed micro-heterogeneous ThO2-UO2 duplex fuel for our proposed marine core in order to explore its capability. For the proposed civil marine propulsion core design, this study uses 18% U-235 enriched micro-heterogeneous ThO2-UO2 duplex fuel. To provide a basis for comparison we also evaluate the performance of homogeneously mixed 15% U-235 enriched all-UO2 fuel. This research also attempts to design a high power density core with 14 EFPY while satisfying the neutronic and thermal-hydraulics safety constraints. A core with an average power density of 100 MW/m3 has been successfully designed while obtaining a core life of 14 years. The average core power density for this core is increased by ∼50% compared to the reference core design (63 MW/m3 and is equivalent to Sizewell B PWR (101.6 MW/m3 which means capital costs could be significantly reduced and the economic attractiveness of the marine core commensurately improved. In addition, similar to the standard SMR core, a reference core with a power density of 63 MW/m3 has been successfully designed while obtaining a core life of ∼16 years. One of the most important points that can be drawn from these studies is that a duplex fuel lattice needs less burnable absorber than uranium-only fuel to achieve the same poison performance. The higher initial reactivity suppression and relatively smaller reactivity swing of the duplex can make the task of reactivity control through BP design in a thorium-rich core easier. It is also apparent that control rods have greater worth in a duplex core, reducing the control material requirements and thus potentially the cost of the rods. This research also analyzed the feasibility of using thorium-based duplex fuel in different cases and environments to observe whether this fuel consistently exhibit superior performance compared to the UO2 core in both the assembly and whole-core levels. The duplex fuel/core consistently exhibits superior performance in consideration of all the neutronic and TH constraints specified. It can therefore be concluded from this study that the superior performance of the thorium-based micro-heterogeneous ThO2-UO2 duplex fuel provides enhanced confidence that this fuel can be reliably used in high power density and long-life SBF marine propulsion core systems, offering neutronic advantages compared to the all-UO2 fuel. Last, but not least, considering all these factors, duplex fuel can potentially open the avenue for low-enriched uranium (LEU) SBF cores with different configurations. Motivated by growing environmental concerns and anticipated economic pressures, the overall goal of this study is to examine the technological feasibility of expanding the use of nuclear propulsion to civilian maritime shipping and to identify and propose promising candidate core designs.

The chronology of the Late Bronze and Early Iron Age Near East is currently a topic of intense scholarly debate. The conventional/orthodox chronology for this period has been assembled over the past one-two centuries using information from King-lists, royal annals and administrative documents, primarily those from the Great Kingdoms of Egypt, Assyria and Babylonia. This major enterprise has resulted in what can best be described as an extremely complex but little understood jigsaw puzzle composed of a multiplicity of loosely connected data. I argue in my thesis that this conventional chronology is fundamentally wrong, and that Egyptian New Kingdom (Memphite) dates should be lowered by 200 years to match historical actuality. This chronological adjustment is achieved in two stages: first, the removal of precisely 85 years of absolute Assyrian chronology from between the reigns of Shalmaneser II and Ashur-dan II; and second, the downward displacement of Egyptian Memphite dates relative to LBA Assyrian chronology by a further 115 years. Moreover, I rely upon Kuhnian epistemology to structure this alternate chronology so as to make it methodologically superior to the conventional chronology in terms of historical accuracy, precision, consistency and testability.