Submitted by Marcele Costal de Castro (costalcastro@gmail.com) on 2017-09-27T15:49:43Z No. of bitstreams: 1 CESAR MARQUES SALGADO D.PDF: 3287236 bytes, checksum: e8cb423520d25b201049a40e5dc0babf (MD5) / Made available in DSpace on 2017-09-27T15:49:43Z (GMT). No. of bitstreams: 1 CESAR MARQUES SALGADO D.PDF: 3287236 bytes, checksum: e8cb423520d25b201049a40e5dc0babf (MD5) Previous issue date: 2010-02 / Este trabalho apresenta uma nova metodologia baseada nos princípios de atenuação de raios gama, por meio de simulações de Monte Carlo (MC), e redes neurais artificiais (RNAs) supervisionadas para predições de frações de volume e identificação de regimes de fluxo em sistemas multifásicos tipo, gás, água e óleo encontrados na indústria petrolífera off-shore. O princípio baseia-se no reconhecimento das distribuições de altura de pulsos obtidas por detectores cintiladores que são utilizadas inteiramente para alimentar de forma simultânea as RNAs. As curvas-resposta (resolução energética e eficiência) de um detector real são consideradas. O sistema de detecção simulado utiliza dois detectores NaI(Tl) e duas energias de raios gama com feixe largo. A geometria proposta considera os feixes transmitido e espalhado tornando o sistema menos dependente do regime de fluxo. O conjunto de padrões necessário para treinamento e avaliação das RNAs foi gerado por meio do código computacional MCNP-X baseado no método de MC a partir de modelos teóricos ideais e estáticos de regimes multifásicos - anular, estratificado e homogêneo. As RNAs mapearam adequadamente os dados simulados com as frações de volume sem a necessidade do conhecimento, a priori, do regime de fluxo. As RNAs identificaram corretamente todos os regimes com predição satisfatória das frações de volume em sistemas multifásicos indicando a possibilidade de aplicação desta metodologia para tal propósito. / This work presents a new methodology for flow regimes identification and volume fractions prediction in gas-water-oil multiphase systems found in off-shore petroleum industry. The approach is based on gamma-ray pulse height distributions (PHDs) pattern recognition by means the artificial neural networks (ANNs). The detection system uses appropriate fan beam geometry, comprised of a dual-energy gamma-ray source and two NaI(Tl) detectors adequately positioned in order calculate transmitted and scattered beams, which makes it less dependent on the flow regime. The system comprises four ANNs, the first identifies the flow regime and the other three ANNs are specialized in volume fraction prediction for each specific regime. The PHDs are directly used by the ANNs without any parameterization of the measured signal. The energy resolution and efficiency of NaI(Tl) detectors are also considered on the mathematical model. The ideal and static theoretical models for annular, stratified and homogenous flow regimes have been developed using MCNP-X mathematical code (simulations by means of Monte Carlo method), which was used to provide training, test and validation data for the ANNs. The proposed ANNs could correctly identified all three different regimes with satisfactory prediction of volume fraction in gas-water-oil multiphase system demonstrating to be a promising approach for this purpose.

Swedish Nuclear Fuel and Waste Management Company- SKB has conducted measurement campaigns at the Swedish central interim storage facility for spent nuclear fuel- CLAB over the years, extending from year 2003 to 2019 where the gamma energy escape was acquired. At CLAB the spent nuclear fuel assembly is inserted into the calorimeter; device intended to measure temperature increase due to decay heat from the fuel assembly. The calorimetric construction is surrounded by water the medium in which the temperature deviations occur and thus are also measured by the calorimeter. However, there is some leakage of gamma energy from the calorimetric construction and does not contribute to the heating of the water. Therefore, only considering the calorimetric measurements is not enough to estimate the total decay heat in the fuel assembly since these measurements fail to account for the gamma escape. Measurements of gamma energy escape acquired over the years at CLAB were observed to have some tendencies that where questionable, mainly some stochastic behavior indicating that their uncertainty was profound. In the scope of the thesis a computational method was developed to calculate the gamma energy escape and thus assist in determining which measurements to discard. Combination of two programs were used one being Spent Nuclear fuel- SNF and the second Monte Carlo N particle Simulator- MCNP, to obtain the computational gamma energy escape for different fuel assemblies and cooling times- CTs. It was established that the escape had a range between 1-3,5% and that it had a dependency on CT, fuel assembly type and operational history. Calculated radial exponential decay coefficient for fuel assemblies of the medium; water had also a clear dependency on CT where values of the coefficient increased over CT. Normalized gamma energy distribution over a rotation around the fuel assembly was calculated and it showed that the assembly tended to have the highest radiation coming from its corner rods. The verification of the computational gamma escape results with corresponding measurements yielded that the agreement was quite good for the earlier measurement campaigns. However, deviation became evident after the 2007 campaign where the calculated values were underestimated compared to the measured.

<p>Elevated blood pressure (BP) is a significant risk factor for cardiovascular diseases (CVD), which are the leading cause of morbidity and mortality. Dietary minerals such as sodium (Na) and potassium (K) play a crucial role in overall health and play a specific function in regulating blood pressure in the human body. Numerous studies have been conducted on the association between blood pressure and dietary intervention. While many nutritional intervention studies have shown adverse effects of excessive Na intake and the beneficial impact of supplemental K in humans, less is understood on Na and K tissue retention and health outcomes of such retention. The most commonly used biomarkers to study Na retention and regulation is urine Na. However, the use of urine Na concentration as an indicator of Na retention has its limitations and has been recently questioned. In-vivo neutron activation analysis (IVNAA) is a unique and powerful technique for elemental analysis in the human body that has the potential to quantify Na and K retention and monitor their bio-kinetics. This research work designed an in vivo neutron irradiation system with high sensitivity and minimal radiation dose to measure Na/K and monitor Na/K bio-kinetics. The system was characterized, tested, and validated for K measurement in mice and rats. Moreover, we developed a methodology for in vivo quantification of Na in pigs in bone and soft tissue after dietary intervention. The project's overall goal is to exploit the potential of a compact DD neutron generator-based neutron activation analysis system for in vivo quantification of Na and K in humans and small animals.</p>

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Dedicated sensors are widely used throughout many industries to monitor everyday operations, maintain safety, and report performance characteristics. In order to adopt a more sustainable solution, much research is being applied to self-powered sensing, implementing solutions which harvest wasted ambient energy sources to power these dedicated sensors. The adoption of not only wireless sensor nodes, but also self-powered capabilities in the nuclear energy process is critical as it can address issues in the overall safety and longevity of nuclear power. The removal of wires for data and power transmission can greatly reduce the cost of both installation and upkeep of power plants, while self-powered capabilities can further reduce effort and money spent in replacing batteries, and importantly may enable sensors to work even in losses to power across the plant, increasing plant safety. This thesis outlines three harvesting opportunities in the nuclear energy process from: thermal, vibration, and radiation sources in the main structure of the power plant, and from thermal and radiation energy from spent fuel in dry cask storage. Thermal energy harvesters for the primary and secondary coolant loops are outlined, and experimental analysis done on their longevity in high-radiation environments is discussed. A vibrational energy harvester for large rotating plant machine vibration is designed, prototyped, and tested, and a model is produced to describe its motion and energy output. Finally, an introduction to the design of a gamma radiation and thermal energy harvester for spent nuclear fuel canisters is discussed, and further research steps are suggested. / Master of Science / In this work multiple energy harvesters are investigated aimed at collecting wasted ambient energy to locally power sensor nodes in nuclear power plants, and in spent nuclear fuel canisters. Locally self-powered, wireless sensors can increase safety and reliability throughout the nuclear process. To address this a thermal energy harvester is tested in a radiation rich environment, and its performance before and after irradiation is analyzed. A vibrational energy harvester designed for use on large rotating machinery is discussed, manufactured, and tested, and a mathematical model describing it is produced. Finally, an introduction to harvesting radiation and heat given off from spent nuclear fuel in dry cask canister storage is investigated. Power capabilities for each design are considered, and the impact of such energy harvesting for wireless sensor nodes on the longevity, safety, and reliability of nuclear power plants is discussed.

<p>Accelerator-driven systems for transmutation of nuclear waste have been suggested as a means for dealing with spent fuel components that pose potential radiological hazard for long periods of time. While not entirely removing the need for underground waste repositories, this nuclear waste incineration technology provides a viable method for reducing both waste volumes and storage times. Potentially, the time spans could be diminished from hundreds of thousand years to merely 1.000 years or even less. A central aspect for accelerator-driven systems design is the prediction of safety parameters and fuel economy. The simulations performed rely heavily on nuclear data and especially on the precision of the neutron cross section representations of essential nuclides over a wide energy range, from the thermal to the fast energy regime. In combination with a more demanding neutron flux distribution as compared with ordinary light-water reactors, the expanded nuclear data energy regime makes exploration of the cross section sensitivity for simulations of accelerator-driven systems a necessity. This fact was observed throughout the work and a significant portion of the study is devoted to investigations of nuclear data related effects. The computer code package EA-MC, based on 3-D Monte Carlo techniques, is the main computational tool employed for the analyses presented. Directly related to the development of the code is the extensive IAEA ADS Benchmark 3.2, and an account of the results of the benchmark exercises as implemented with EA-MC is given. CERN's Energy Amplifier prototype is studied from the perspectives of neutron source types, nuclear data sensitivity and transmutation. The commissioning of the n_TOF experiment, which is a neutron cross section measurement project at CERN, is also described.</p>

Accelerator-driven systems for transmutation of nuclear waste have been suggested as a means for dealing with spent fuel components that pose potential radiological hazard for long periods of time. While not entirely removing the need for underground waste repositories, this nuclear waste incineration technology provides a viable method for reducing both waste volumes and storage times. Potentially, the time spans could be diminished from hundreds of thousand years to merely 1.000 years or even less. A central aspect for accelerator-driven systems design is the prediction of safety parameters and fuel economy. The simulations performed rely heavily on nuclear data and especially on the precision of the neutron cross section representations of essential nuclides over a wide energy range, from the thermal to the fast energy regime. In combination with a more demanding neutron flux distribution as compared with ordinary light-water reactors, the expanded nuclear data energy regime makes exploration of the cross section sensitivity for simulations of accelerator-driven systems a necessity. This fact was observed throughout the work and a significant portion of the study is devoted to investigations of nuclear data related effects. The computer code package EA-MC, based on 3-D Monte Carlo techniques, is the main computational tool employed for the analyses presented. Directly related to the development of the code is the extensive IAEA ADS Benchmark 3.2, and an account of the results of the benchmark exercises as implemented with EA-MC is given. CERN's Energy Amplifier prototype is studied from the perspectives of neutron source types, nuclear data sensitivity and transmutation. The commissioning of the n_TOF experiment, which is a neutron cross section measurement project at CERN, is also described.

Les activités nucléaires génèrent des déchets radioactifs classés selon leur niveau d’activité et la durée de vie des radioéléments présents. La garantie d’un classement et d’une gestion optimale nécessite une caractérisation précise. Les déchets de moyenne et haute activité, contenant des radioéléments à vie très longue, seront stockés en profondeur pendant plusieurs centaines de milliers d’années, à l’issue desquelles il est nécessaire de pouvoir garantir l’absence de risques pour l’homme et l’environnement, non seulement sur le plan radiologique, mais aussi en ce qui concerne des éléments stables, toxiques du point de vue chimique. Cette thèse concerne la caractérisation par activation neutronique de ces éléments toxiques, ainsi que celle des matières nucléaires présentes dans les colis. Elle a été réalisée dans le cadre d’une collaboration entre le Laboratoire de Mesures Nucléaires du CEA Cadarache, en France, et l’institut de Gestion des Déchets Radioactifs et de Sûreté des Réacteurs du centre de recherche FZJ (Forschungszentrum Jülich), en Allemagne. La première étude a consisté à valider le modèle numérique de la cellule d’activation neutronique MEDINA (FZJ) avec le code de transport Monte Carlo MCNP. Les rayonnements gamma prompts de capture radiative d’échantillons contenant des éléments d’intérêt (béryllium, aluminium, chlore, cuivre, sélénium, strontium et tantale) ont été mesurés et comparés aux simulations avec diverses bases de données nucléaires, permettant d’aboutir à un accord satisfaisant et validant le schéma de calcul en vue des études suivantes. Ensuite, la mesure des rayonnements gamma retardés de fissions induites sur les isotopes 235U et 239Pu a été étudiée pour des fûts de 225 L contenant des enrobés bitumineux ou une matrice béton, représentatifs de déchets produits en France et en Allemagne. Les rendements d’émission des rayonnements gamma retardés de fission d’intérêt, cohérents avec ceux publiés dans la littérature, ont été déterminés à partir des mesures d’échantillons métalliques d’uranium et de plutonium dans la cellule d’activation neutronique REGAIN du LMN. Le signal utile a ensuite été extrapolé par simulation MCNP pour une répartition homogène d’isotopes 239Pu ou 235U dans les matrices considérées, en utilisant le modèle numérique de MEDINA. Des signaux faibles, de l’ordre de 100 coups par gramme d’isotope 239Pu ou 235U, ont été obtenus. Pour le colis d’enrobés bitumineux, le niveau d’irradiation gamma très élevé, dû à une activité en 137Cs de l’ordre de 1 TBq par fût, nécessiterait l’utilisation d’une collimation et/ou d’écrans pour éviter la saturation de l’électronique de mesure, rendant indétectables les rayonnements gamma retardés de fission. Les colis de déchets bétonnés produits en Allemagne présentant un niveau d’activité plus faible, il a été possible d’estimer des limites de détection allant de 10 à 290 g d’isotope fissile 235U ou 239Pu, selon la raie gamma considérée, suite à la mesure du bruit de fond actif dans MEDINA avec une matrice béton maquette. Afin d’améliorer ces performances, le blindage du détecteur germanium de MEDINA a été optimisé à l’aide de simulations MCNP, montrant la possibilité de réduire les bruits de fond gamma et neutron d’un facteur 4 et 5, respectivement. La validation expérimentale de l’efficacité du blindage a été effectuée à partir de configuration simples à implémenter dans MEDINA, confirmant les facteurs de réduction attendus. Un blindage du détecteur optimal permettrait d’améliorer les limites de détection et aussi d’utiliser une source de neutrons d’intensité supérieure, comme un générateur de neutron à haut flux ou un accélérateur linéaire d’électrons avec une cible de conversion appropriée. / Nuclear activities produce radioactive wastes classified following their radioactive level and decay time. An accurate characterization is necessary for efficient classification and management. Medium and high level wastes containing long lived radioactive isotopes will be stored in deep geological storage for hundreds of thousands years. At the end of this period, it is essential to ensure that the wastes do not represent any risk for humans and environment, not only from radioactive point of view, but also from stable toxic chemicals. This PhD thesis concerns the characterization of toxic chemicals and nuclear material in radioactive waste, by using neutron activation analysis, in the frame of collaboration between the Nuclear Measurement Laboratory of CEA Cadarache, France, and the Institute of Nuclear Waste Management and Reactor Safety of the research center, FZJ (Forschungszentrum Jülich GmbH), Germany. The first study is about the validation of the numerical model of the neutron activation cell MEDINA (FZJ), using MCNP Monte Carlo transport code. Simulations and measurements of prompt capture gamma rays from small samples measured in MEDINA have been compared for a number of elements of interest (beryllium, aluminum, chlorine, copper, selenium, strontium, and tantalum). The comparison was performed using different nuclear databases, resulting in satisfactory agreement and validating simulation in view of following studies. Then, the feasibility of fission delayed gamma-ray measurements of 239Pu and 235U in 225 L waste drums has been studied, considering bituminized or concrete matrixes representative of wastes produced in France and Germany. The delayed gamma emission yields were first determined from uranium and plutonium metallic samples measurements in REGAIN, the neutron activation cell of LMN, showing satisfactory consistency with published data. The useful delayed gamma signals of 239Pu and 235U, homogeneously distributed in the 225 L matrixes, were then determined by MCNP simulations using MEDINA numerical model. Weak signals of about one hundred counts per gram of 239Pu or 235U after 7200 s irradiation were obtained. Because of the high gamma emission in the bituminized waste produced in France (about 1 TBq of 137Cs per drum), the use of collimator and/or shielding is mandatory to avoid electronic saturation, making fission delayed gamma rays undetectable. However, German concrete drums being of lower activity, their corresponding active background was measured in MEDINA with a concrete mock-up, leading to detection limits between 10 and 290 g of 235U or239Pu, depending on the delayed gamma line. In order to improve these performances, the shielding of MEDINA germanium detector was optimized using MCNP calculations, resulting in gamma and neutron background reduction factors of 4 and 5, respectively. The experimental validation of the shielding efficiency was performed by implementing easy-to-build configurations in MEDINA, which confirmed the expected background reduction factors predicted by MCNP. Thanks to an optimized detector shielding, it will also be possible to use a higher neutron emission source, like a high flux neutron generator or an electron LINAC with appropriate conversion targets, in view to further reduce detection limits.

Aujourd'hui, nous sommes tous témoins d'une compétition des installations en pays différents pour étudier les régions inconnues de noyaux riches en neutrons. Beaucoup d'efforts sont consacrés à comprendre le rôle de l'excès de neutrons et son influence sur les noyaux dans les environs de coquilles de neutrons fermées. Un des moyens pour étudier la structure nucléaire est via la désintégration bêta. Une fois un noyau est prouvé d'exister, ses propriétés de désintégration bêta, comme T1/2 et Pn (probabilité de l'émission de neutrons de bêta-retardés), qui sont relativement faciles à mesurer, peuvent fournir les premiers indices sur la structure nucléaire. Sur le site de processus-r des «points d'attente» (noyaux sur des coquilles de neutrons fermés) ont des effets importants sur la dynamique processus-r ainsi que sur la distribution de l'abondance des éléments. Les conditions astrophysiques exactes en vertu de desquelles la synthèse nucléaire a lieu ne sont pas connus avec certitude. Parce que les noyaux participant en processus-r sont difficiles à synthétiser même aujourd'hui et à étudier expérimentalement, les paramètres nécessaires pour les calculs du processus-r sont principalement dérivés de modèles théoriques. Comme on l'a vu récemment, la plupart des théories n'ont pas réussi à reproduire des ensembles de données nouvellement mesurées près de fermetures de couche. Avec de nouvelles données expérimentales déjà (ou bientôt) disponibles les approches théoriques peuvent être ajustées. Comme émission de neutrons retardée bêta devient plus importante voie le canal dominant en désintégration des noyaux loin d'un stabilité, l'utilisation d'un détecteur de neutrons approprié afin d'étudier leurs propriétés est indispensable. C'est pour mener la recherche appropriée que dans le cadre de la thèse, en étroite collaboration avec le JINR (Dubna) un nouveau système de détection a été construit. Il se compose de 80 compteurs de ³He, un détecteur 4π de bêta et un HPGe afin de mesurer simultanément l'activité de gamma, bêta et neutrons. Le développement d'un tel système de détection, actuellement installé sur ALTO, a été le premier objectif de la thèse. Puis, au cours de deux campagnes expérimentales menées pour examiner les propriétés de désintégration bêta de noyaux riches en neutrons dans le proximité de N = 50, N = 82 la capacité de travail du système de détection produit a été prouvée. Dans le région de ⁷⁸Ni : le temps de vie et de la probabilité de l'émission de neutrons bêta retardés pour ⁸º,⁸²,⁸³,⁸⁴Ga ont été mesurés. Nous sommes les premiers à observer la structure de ⁸¹,⁸² Ge via spectroscopie gamma spectre conditionnée par bêta et par neutron. Grâce à la détection des neutrons les intensités absolues de la désintégration bêta ont été proposées pour la première fois. Dans le région de ¹³²Sn les temps de vie de ¹²³Ag, ¹²⁴Ag, ¹²⁵Ag et ¹²⁷In, ¹²⁸In ont été mesurées. Pour la première fois l'émission de neutrons retardés bêta a été observée pour le cas de ¹²⁶Cd, sa valeur Pn également mesurée. Sur la base des données obtenues, nous arrivons à la conclusion que, pour déterminer la contribution relative de désintégrations permises et interdites les efforts théoriques doivent être faites en traversant la couche N = 50. Alors que dans le région de N = 82 plus de données expérimentales sont nécessaires. / Nowadays we are all witnesses of a competition of facilities at different countries to study unknown regions of neutron rich nuclei. Much efforts are devoted to understand the role of neutron excess and its influence on nuclei in vicinity of closed neutron shells. One of the means to investigate nuclear structure is in beta-decay. Once a nucleus is proven to exist, its beta-decay properties, such as T1/2 and Pn (probability of beta-delayed neutron emission), which are relatively easy to measure, can provide the first hints on the nuclear structure. On the r-process site, "waiting points"(nuclei on closed neutron shells) has significant effects on the r-process dynamics and the abundance distribution. The actual side and the astrophysical conditions under which the nuclear synthesis takes place are still not certainly known - since r-process nuclei are difficult to produce and to study experimentally, input parameters for r-process calculations are mostly derived from theoretical models. As it has been seen lately, most of the theories have failed to reproduce newly measured data sets near shell closures. With new experimental data already (or shortly) available theoretical approaches can be adjusted. Since a beta-delayed neutron emission becomes strong if not dominating decaying channel for nuclei far stability, a proper neutron detector to study their properties is indispensable. To conduct the appropriate investigations, in the frame of the present thesis, in close collaboration with JINR (Dubna) a new detection system was constructed. It consists of 80 ³He-filled counters, 4π beta detector and a HPGe in order to measure simultaneously beta, gamma, neutron activity. The development of such a detection system system, currently installed at ALTO ISOL facility, was the first objective of the thesis. Then, during two experimental campaigns conducted to investigate beta decay properties of neutron rich nuclei in the neighborhood of N=50, N=82 the workability of the newly produced detection system was proven. In the vicinity of ⁷⁸Ni: half- lives and probability of beta-delayed neutron emission for ⁸º,⁸²,⁸³,⁸⁴Ga were measured. We were the first to observe the structure of ⁸¹,⁸² Ge via beta neutron gated gamma spectra. Thanks to the neutron detection channel the absolute intensities of beta decay were proposed for the first time. In the vicinity of ¹³²Sn the half lives of ¹²³Ag, ¹²⁴Ag, ¹²⁵Ag and ¹²⁷In, ¹²⁸In was measured. For the first time the beta delayed neutron emission was observed for ¹²⁶Cd, its Pn value also measured. Based on the data obtained we come to the conclusion that to figure out the relative contribution of allowed and forbidden decays more theoretical efforts should be done crossing the N=50 shell. Whereas in the vicinity of N=82 shell more experimental challenge are required.