Le dioxyde d'uranium (UO2) est le combustible nucléaire le plus largement répandu dans le monde pour alimenter les centrales nucléaires et plus particulièrement les réacteurs à eau pressurisée (REP). En réacteur, la fission des atomes d'uranium crée des produits de fission et des défauts ponctuels dans le matériau combustible. La compréhension de l'évolution de ces dégâts d'irradiation nécessite une approche de modélisation multi-échelle, de l'échelle de la pastille combustible à l'échelle atomique. Nous avons utilisé une méthode de calcul de structure électronique (DFT), pour modéliser les dégâts d'irradiation dans UO2 à l'échelle atomique. Un terme d'interaction Coulombienne de type Hubbard est ajouté au formalisme de la DFT standard pour prendre en compte les fortes corrélations des électrons 5f dans l'UO2. Cette méthode a été utilisée pour étudier les défauts ponctuels dans différents états de charge ainsi que l'incorporation et la diffusion du krypton dans le dioxyde d'uranium. Cette étude nous a permis d'obtenir des données clés pour les modèles aux échelles supérieures mais aussi pour interpréter des résultats expérimentaux. En parallèle de cette étude, trois pistes d'amélioration de l'état de l'art des calculs pour la description de l'UO2 ont été explorées : la prise en compte du couplage spin-orbite, l'application de fonctionnelles permettant la prise en compte des interactions non locales telles que les interactions de van der Waals importantes pour les gaz rares et l'utilisation de la théorie de champ dynamique moyen (Dynamical Mean Field Theory) combinée à la DFT afin de prendre en compte les corrélations dynamiques des électrons 5f. / Uranium dioxide (UO2) is worldwide the most widely used fuel in nuclear plants in the world and in particular in pressurized water reactors (PWR). In-pile the fission of uranium nuclei creates fission products and point defects in the fuel. The understanding of the evolution of these radiation damages requires a multi-scale modelling approach of the nuclear fuel, from the scale of the pellet to the atomic scale. We used an electronic structure calculation method based on the density functional theory (DFT) to model radiation damage in UO2 at the atomic scale. A Hubbard-type Coulomb interaction term is added to the standard DFT formalism to take into account the strong correlations of the 5f electrons in UO2. This method is used to study point defects with various charge states and the incorporation and diffusion of krypton in uranium dioxide. This study allowed us to obtain essential data for higher scale models but also to interpret experimental results. In parallel of this study, three ways to improve the state of the art of electronic structure calculations of UO2 have been explored: the consideration of the spin-orbit coupling neglected in current point defect calculations, the application of functionals allowing one to take into account the non-local interactions such as van der Waals interactions important for rare gases and the use of the Dynamical Mean Field Theory combined to the DFT method in order to take into account the dynamical effects in the 5f electron correlations.

A análise por ativação com nêutrons é um método utilizado na determinação de diversos elementos em diferentes tipos de matrizes. Entretanto, quando a amostra contém altos teores de U ocorre o problema de interferência devido aos produtos de fissão do isótopo 235U. Um dos métodos de tratar este problema é fazer a correção usando fatores de interferência devido à fissão do U para os radionuclídeos utilizados nas análises dos elementos. No presente estudo foram determinados os valores dos fatores de interferência devido à fissão do U para os radioisótopos 141Ce, 143Ce,140La, 99Mo, 147Nd, 153Sm e 95Zr no reator nuclear de pesquisas IEA-R1 do IPEN-CNEN/SP. Esses fatores de interferência foram determinados experimentalmente, por meio da irradiação dos padrões sintéticos em uma determinada posição do reator, e teoricamente, determinando a razão dos fluxos de nêutrons epitérmicos e térmicos na mesma posição onde os padrões sintéticos foram irradiados e utilizando parâmetros nucleares da literatura. Os fatores de interferência obtidos foram comparados com os valores reportados em outros estudos. Para avaliar esses fatores de interferência, eles foram aplicados em análises dos elementos alvo deste estudo, nos materiais de referência certificados NIST 8704 Buffalo River Sediment, IRMM BCR-667 Estuarine Sediment e IAEA-SL-1 Lake Sediment. / Neutron activation analysis is a method used in the determination of several elements in different kinds of matrices. However, when the sample contains high U levels the problem of 235U fission interference occurs. A way to solve this problem is to perform the correction using the interference factor due to U fission for the radionuclides used on elemental analysis. In this study, the interference factors due to U fission for the radioisotopes 141Ce, 143Ce, 140La, 99Mo, 147Nd, 153Sm and 95Zr in the research nuclear reactor IEA-R1 at IPEN-CNEN/SP were determined. These interference factors were determined experimentally, by irradiation of synthetic standards in a selected position in the reactor, and theoretically, determining the epithermal to neutron fluxes ratio in the same position where synthetic standards were irradiated and using reported nuclear parameters on the literature. The obtained interference factors were compared with values reported by other works. To evaluate the reliability of these factors they were applied in the analysis of studied elements in the certified reference materials NIST 8704 Buffalo River Sediment, IRMM BCR- 667 Estuarine Sediment e IAEA-SL-1 Lake Sediment.

A análise por ativação com nêutrons é um método utilizado na determinação de diversos elementos em diferentes tipos de matrizes. Entretanto, quando a amostra contém altos teores de U ocorre o problema de interferência devido aos produtos de fissão do isótopo 235U. Um dos métodos de tratar este problema é fazer a correção usando fatores de interferência devido à fissão do U para os radionuclídeos utilizados nas análises dos elementos. No presente estudo foram determinados os valores dos fatores de interferência devido à fissão do U para os radioisótopos 141Ce, 143Ce,140La, 99Mo, 147Nd, 153Sm e 95Zr no reator nuclear de pesquisas IEA-R1 do IPEN-CNEN/SP. Esses fatores de interferência foram determinados experimentalmente, por meio da irradiação dos padrões sintéticos em uma determinada posição do reator, e teoricamente, determinando a razão dos fluxos de nêutrons epitérmicos e térmicos na mesma posição onde os padrões sintéticos foram irradiados e utilizando parâmetros nucleares da literatura. Os fatores de interferência obtidos foram comparados com os valores reportados em outros estudos. Para avaliar esses fatores de interferência, eles foram aplicados em análises dos elementos alvo deste estudo, nos materiais de referência certificados NIST 8704 Buffalo River Sediment, IRMM BCR-667 Estuarine Sediment e IAEA-SL-1 Lake Sediment. / Neutron activation analysis is a method used in the determination of several elements in different kinds of matrices. However, when the sample contains high U levels the problem of 235U fission interference occurs. A way to solve this problem is to perform the correction using the interference factor due to U fission for the radionuclides used on elemental analysis. In this study, the interference factors due to U fission for the radioisotopes 141Ce, 143Ce, 140La, 99Mo, 147Nd, 153Sm and 95Zr in the research nuclear reactor IEA-R1 at IPEN-CNEN/SP were determined. These interference factors were determined experimentally, by irradiation of synthetic standards in a selected position in the reactor, and theoretically, determining the epithermal to neutron fluxes ratio in the same position where synthetic standards were irradiated and using reported nuclear parameters on the literature. The obtained interference factors were compared with values reported by other works. To evaluate the reliability of these factors they were applied in the analysis of studied elements in the certified reference materials NIST 8704 Buffalo River Sediment, IRMM BCR- 667 Estuarine Sediment e IAEA-SL-1 Lake Sediment.

The closure of the nuclear fuel cycle is a topic of interest in the sustainability context of nuclear energy. The implication of such closure includes considerations of nuclear waste management. This originates from the fact that a closed fuel cycle requires recycling of useful materials from spent nuclear fuel and discarding of non-usable streams of the spent fuel, which are predominantly the fission products. The fission products represent the near-term concerns associated with final geological repositories for the waste stream. Long-lived fission products also contribute to the long-term concerns associated with such repository. In addition, an ultimately closed nuclear fuel cycle in which all actinides from spent nuclear fuels are incinerated will result in fission products being the only source of radiotoxicity. Hence, it is desired to develop a transmutation strategy that will achieve reduction in the inventory and radiological parameters of significant fission products within a reasonably short time. In this dissertation, a transmutation strategy involving the use of the VHTR is developed. A set of specialized metrics is developed and applied to evaluate performance characteristics. The transmutation strategy considers six major fission products: 90Sr, 93Zr, 99Tc, 129I, 135Cs and 137Cs. In this approach, the unique core features of VHTRs operating in equilibrium fuel cycle mode of 405 effective full power days are used for transmutation of the selected fission products. A 30 year irradiation period with 10 post-irradiation cooling is assumed. The strategy assumes no separation of each nuclide from its corresponding material stream in the VHTR fuel cycle. The optimum locations in the VHTR core cavity leading to maximized transmutation of each selected nuclides are determined. The fission product transmutation scenarios are simulated with MCNP and ORIGEN-S. The results indicate that the developed fission product transmutation strategy offers an excellent potential approach for the reduction of inventories and radiological parameters, particularly for long-lived fission products (93Zr, 99Tc, 129I and 135Cs). It has been determined that the in-core transmutation of relatively short-lived fission products (90Sr and 137Cs) has minimal advantage over a decay-only scenario for these nuclides. It is concluded that the developed strategy is a viable option for the reduction of radiotoxicity contributions of the selected fission products prior to their final disposal in a geological repository. Even in the cases where the transmutation advantage is minimal, it is deemed that the improvement gained, coupled with the virtual storage provided for the fission products during the irradiation period, makes the developed fission product transmutation strategy advantageous in the spent fuel management scenarios. Combined with the in-core incineration options for TRU, the developed transmutation strategy leads to potential achievability of engineering time scales in the comprehensive nuclear waste management.

A high-efficiency gamma-ray counting system has been built at Los Alamos National Laboratory for use in analyzing nuclear forensics samples. This system consists of two clover high-purity germanium detectors and is surrounded by a thallium-doped sodium iodide annulus. Special precautions have been taken to ensure the system has a low background. The system is connected to XIA Pixie-4 fast digitizers and collects data in list-mode. This work is split into two main parts. The first part describes the proper steps and techniques to initialize the settings of a detector system connected to fast digitizers in order to optimize the system for resolution and throughput. The various counting modes for this particular system are described in detail, including the benefits and drawbacks of each mode. Steps are then shown to characterize the system by obtaining efficiency curves for various counting modes and sample geometries. Because of the close counting geometry involved with this system, true-coincidence summing factors must be calculated, and are done so in part by measuring the peak-to-total ratios of the system in its various counting modes across a wide energy range. The dead-time for the system can be complicated due to the multiple inputs of the system. Techniques for calculating the dead-time of multiple-detector systems are discussed. The second part of this work shows the system's usefulness in analyzing nuclear forensics samples, specifically irradiated enriched uranium. Three fission product parent-daughter pairs of different lifetimes are analyzed over a course of six months. The activities of each nuclide are calculated at each time step. Age dating techniques using the parent-daughter pairs are discussed, as well as the detection limits of each nuclide for a range of sample ages. Finally, avenues for further research are presented, as well as potential sources of error or uncertainty for this work. / text

Made available in DSpace on 2014-10-09T12:27:01Z (GMT). No. of bitstreams: 0 / Made available in DSpace on 2014-10-09T13:56:47Z (GMT). No. of bitstreams: 0 / Fundação de Amparo à Pesquisa do Estado de São Paulo (FAPESP) / Tese (Doutoramento) / IPEN/T / Instituto de Pesquisas Energeticas e Nucleares - IPEN-CNEN/SP / FAPESP:06/54851-8

Made available in DSpace on 2014-10-09T12:33:04Z (GMT). No. of bitstreams: 0 / Made available in DSpace on 2014-10-09T14:06:06Z (GMT). No. of bitstreams: 0 / Dissertação (Mestrado) / IPEN/D / Instituto de Pesquisas Energeticas e Nucleares - IPEN-CNEN/SP

Made available in DSpace on 2014-10-09T12:27:01Z (GMT). No. of bitstreams: 0 / Made available in DSpace on 2014-10-09T13:56:47Z (GMT). No. of bitstreams: 0 / Fundação de Amparo à Pesquisa do Estado de São Paulo (FAPESP) / Uma média de 80 % dos radiofármacos usados nas clínicas são marcados com 99mTc por suas propriedades físicas adequadas e fácil obtenção através de geradores de 99Mo/99mTc. A Diretoria de Radiofarmácia (DIRF) do IPEN-CNEN/SP desenvolveu um gerador cromatográfico tipo gel de MoZr com 99Mo produzido pela da reação 98Mo(n,)99Mo que ocorre no reator Nuclear IEA-R1 do IPEN-CNEN/SP. O gel é composto de molibdato de zircônio com volume de eluição de 12 mL com uma atividade de 11100 MBq (300 mCi) produzindo uma concentração radioativa de 925 MBq (25 mCi)/mL. O gerador de fissão produz uma concentração radioativa maior, de 1850 MBq (50 mCi)/mL. Pretende-se com esse trabalho desenvolver um gerador 99 Mo/99mTc tipo gel com a qual se possa eluir 99mTc obtendo-se uma concentração radioativa adequada para atender a demanda de mercado sem perder a qualidade. Foram desenvolvidos dois tipos de sistemas de concentração o único e o em série. O sistema mais adequado para o gerador de 99Mo/99mTc do tipo gel de MoZr estéril e automatizado à vácuo foi o sistema de concentração em série utilizando o cartucho Dionex 2,5 cc/QMA. O gerador de gel é eluído com 10 mL de solução de NaCl 0,1 % sendo o pertecnetato retido no cartucho aniônico QMA e eluído com 4 mL de solução de NaCl de 0,9 %. O processo dura no máximo 30 minutos. A eficiência de eluição do sistema de concentração foi de 90 %. No início de 2009 aconteceu uma crise mundial do abastecimento de 99Mo fazendo com que surgisse a necessidade do desenvolvimento de tecnologias alternativas para a produção de geradores de 99Mo/99mTc utilizando 99Mo produzido por fissão ou o desenvolvimento de um método adequado para estender a vida útil deste gerador. Os resultados deste trabalho mostraram que é possível utilizar o mesmo sistema de concentração desenvolvido para o gerador de gel, o que levará a um fator de concentração de 3 para o 99mTc eluído. / Tese (Doutoramento) / IPEN/T / Instituto de Pesquisas Energeticas e Nucleares - IPEN-CNEN/SP / FAPESP:06/54851-8

Made available in DSpace on 2014-10-09T12:33:04Z (GMT). No. of bitstreams: 0 / Made available in DSpace on 2014-10-09T14:06:06Z (GMT). No. of bitstreams: 0 / O carvão ativado é um tipo comum de rejeito radioativo que contém elevada concentração de produtos de ativação e fissão. O gerenciamento deste rejeito inclui a sua caracterização, visando à determinação e quantificação dos radionuclídeos específicos, incluindo aqueles conhecidos como Radionuclídeos de Difícil Medição (RDM). A análise dos RDMs geralmente envolve análises radioquímicas complexas para purificação e separação dos radionuclídeos, as quais são caras e demandam muito tempo. O objetivo deste trabalho foi definir uma metodologia de análise sequencial de isótopos de urânio, netúnio, plutônio, amerício e cúrio, presentes em um tipo de rejeito radioativo, avaliando-se rendimento químico, tempo de análise, quantidade de rejeito secundário gerado e custo. Foram comparadas e validadas três metodologias que empregam a troca iônica (TI + EC), extração cromatográfica (EC) e extração com polímeros (ECP). O rejeito estudado foi o carvão ativado, proveniente do sistema de purificação de água do circuito primário de refrigeração do reator IEA-R1. As amostras de carvão foram dissolvidas por digestão ácida, seguida de purificação e separação dos isótopos com resinas de troca iônica, extração cromatográfica e extração com polímeros. Os isótopos foram analisados em um espectrômetro alfa, equipado com detectores de barreira de superfície. O rendimento químico de todos os elementos foi satisfatório para os métodos TI + EC e EC. Para o método ECP, apenas o rendimento químico do U foi comparável aos outros métodos. As análises estatísticas dos resultados bem como a análise de custo e volume de rejeito secundário gerado demonstraram que o método EC é o mais adequado para a identificação e quantificação dos isótopos estudados em carvão ativado. / Dissertação (Mestrado) / IPEN/D / Instituto de Pesquisas Energeticas e Nucleares - IPEN-CNEN/SP

The general fission process is well known and is applied in nuclear power plants all over the world. However many properties of fission fragments are still not well understood. The angular momentum distribution of fission fragments is an important property to gain a better understanding of the fission process, and that can be derived indirectly from isomeric yield ratios. The goal of this project has been to develop a script in Python that runs the nuclear reaction code TALYS with the Total Monte Carlo method to calculate the isomeric yield ratio. The script generates a matrix consisting of excitation energies and angular momenta that is provided to TALYS. One matrix corresponds to one calculation of the isomeric ratio. Thus, the dependency of the isomeric yield ratio on these matrices can be observed. After looking into the matrices, the dependencies of the isomeric yield ratios on the excitation energies and the angular momentum distribution are observed. In this project, the calculated isomeric yield ratios are compared with the experimental value obtained from an experiment conducted in August of 2019 at the IGISOL-JYFLTRAP facility in Jyväskylä, Finland. It is worth mentioning that, fission system is of Uranium-238 which was induced by a proton beam at an energy of 25 MeV. The dependency of the isomeric yield ratio (IYR) on the angular momentum and the excitation energy has been investigated. However, it has proved more difficult than expected, to deduce an estimation for the angular momentum distribution. Another finding of this project is that the two codes used, GEF and TALYS sometimes produce inconstant results.