The Integral Inherently Safe Light Water Reactor (I2S-LWR) is a novel reactor concept which aims to apply safety-promoting features typical of small modular reactors (SMRs) to a large pressurized water reactor (PWR) of 3000 MWt, thus providing an option for a passively safe reactor to markets which would find greater economic benefit in a large reactor. Pushing the compact core of an integral reactor to 3000 MWt necessitates several design innovations to remain within safety margins while meeting the goal of increased power density. The I2S-LWR fuel assembly takes on a 19x19 lattice with reduced fuel rod dimensions relative to traditional Westinghouse-type 17x17 PWR fuel assemblies. It is anticipated that the I2S-LWR will eventually employ uranium silicide (U3Si2) fuel instead of uranium oxide (UO2) to improve thermal performance. These unique design features are closely tied to the I2S-LWR core neutronics, thereby necessitating a thorough investigation of reactivity control options. This thesis considers the design of both control rods and burnable absorbers on the basis of the I2S-LWR uranium silicide fuel assembly. Fuel assembly designs are considered with various control rod arrangements and burnable absorber layouts with several candidate absorber materials and concentrations. Viable fuel assembly designs must meet targets for reactivity and power peaking while satisfying constraints on core safety and cycle length. Designs are developed in a heuristic manner, and key performance metrics are processed at each iteration. Characteristics of common optimization algorithms are mimicked at a high level so as to guide the progression of design iterations. The optimized fuel assembly designs produced in this way are recommended for use in core loading pattern design.

Investigation of nitride fuels and their properties has, for decades, been propelled on the basis of their desirable high metal densities and high thermal conductivities, both of which oer intrinsic advantages to performance, economy, and safety in fast and light water reactor systems. In this time several key obstacles have been identied as impeding the implementation of these fuels for commercial applications; namely chemical interactions with air and steam, the noted diculty in sintering of the material, and the high costs associated with the enrichment of 15N. The combination of these limitations, historically, led to the well founded conclusion that the most appropriate use of nitride fuels was in the fast reactor fuel cycle, where the cost burdens associated with them is substantially less. Indeed, it is within this context that the vast majority of work on nitrides has been and continues to be done. Nevertheless, following the 2011 Fukushima-Daiichi nuclear accident, a concerted governmental-industrial eort was embarked upon to explore the alternatives of so-called \accident tolerant" and \high performance" fuels. These fuels would, at the same time, improve the response of the fuel-clad system to severe accidents and improve the economy of operation for light water reactor systems. Among the various candidates proposed are uranium nitride, uranium silicide, and a third \uranium nitride-silicide" composite featuring a mixture of the former. In this thesis a method has been established for the synthesis, fabrication, and characterization of high purity uranium nitride, and uranium nitride-silicide composites, prepared by the spark plasma sintering (SPS) technique. A specic result has been to isolate the impact of the processing parameters on the microstructure of representative fuel pellets, essentially permitting any conceivable microstructure of interest to be fabricated. This has enabled the development of a highly reproducible technique for the production of pellets with microstructures tailored towards any desired porosity between 88-99.9%TD, any grain size between 6-24 μm, and, in the case of  the uranium nitride-silicide composite, a silicide-coated UN matrix. This has permitted the evaluation of these microstructural characteristics on the performance of these materials, specically with respect to their role as accident tolerant fuels. This has generated results which have tightly coupled nitride performance with pellet microstructure, with important implications for the use of nitrides in water-cooled reactors. / Under artionden har forskning om nitridbranseln och dess egenskaper bedrivits pa grundval av nitridbransletsatravarda egenskaper avseende dess hoga metall tathet och hog varmeledningsformaga. Dessa egenskaper besitter vasentliga fordelar avseende prestanda, ekonomi och sakerhet for metallkylda som lattvatten reaktorer. Genom forskning har aven centrala begr ansningar identierats for implementering av nitridbranslen for kommersiellt bruk. Begransningar avser den kemiska interaktionen med luft och vattenanga, en uppmarksammad svarighet att sintring av materialet samt hoga kostnader forknippade med den nodvandiga anrikningen av 15-N. Kombinationen av dessa begransningar resulterade, tidigare, i en valgrundad slutsats att nitridbranslet mest andamalsenliga anvandningsomrade var i karnbranslecykeln for snabba reaktorer. Detta da kostnaderna forenade med implementeringen av branslet ar avsevart lagre. Inom detta sammanhang har majoriteten av forskning avseende nitrider bedrivits och fortskrider an idag. Dock, efter karnkraftsolyckan i Fukushima-Daiichi 2011, inleddes en samlad industriell och statlig anstrangning for att undersoka alternativ till sa kallade \olyckstoleranta" och \hogpresterande" branslen. Dessa branslen skulle samtidigt forbattra reaktionstiden for bransleinkapsling systemet mot allvarliga olyckor samt forbattra driftsekonomin av lattvattenreaktorer. Foreslagna kandidater ar urannitrid, uransilicid och en tredje \uran nitrid-silicid", komposit bestaende av en blandning av de foregaende. Genom denna avhandling har en metod faststallts for syntes, tillverkning och karaktarisering av uran nitrid av hog renhet samt uran nitrid-silicid kompositer, forberedda med tekniken SPS (Spark Plasma Sintering). Ett specikt resultat har varit att isolera eekten av processparametrar pa mikrostrukturen pa representativa branslekutsar. Detta mojliggor, i princip, framstallningen av alla tankbara mikrostrukturer utav intresse for tillverkning. Vidare har detta mojliggjort utvecklingen av en hogeligen reproducerbar  teknik for framstallningen av branslekutsar med mikrostrukturer skraddarsydda for onskad porositet mellan 88 och 99.9 % TD, och kornstorlek mellan 6 och 24 μm. Dartill har en metod for att belagga en matris av uran nitrid-silicid framarbetats. Detta har mojliggjort utvarderingen av dessa mikrostrukturella parametrars paverkan pa materialens prestanda, sarskilt avseende dess roll som olyckstoleranta branslen. Detta har genererat resultat som ar tatt sammanlankat nitridbranslets prestanda till kutsens mikrostruktur, med viktiga konsekvenser for den potentiella anvandningen av nitrider i lattvatten reaktorer. / <p>QC 20170210</p>

A segurança da operação de reatores nucleares depende dos materiais envolvidos em sua construção, pois são submetidos a variações de temperaturas em ambiente corrosivo e avarias causadas por partículas de alta energia. O combustível, que proporciona energia para o reator, possui vida útil muito menor, mas é submetido às mesmas condições. Dentre as ligas de urânio, o U3Si2 é bastante utilizado em reatores de pesquisa, dada a elevada densidade de urânio, boa condutividade térmica e resistência à amorfização induzida por radiação, ao inchamento e à propagação de trincas. Porém, no processo de fabricação da liga U-Si geralmente são formadas duas ou mais fases cristalinas, com comportamentos distintos sob irradiação. Por esse motivo, a especificação do pó de siliceto de urânio utilizado no reator IEA-R1 do IPEN, e do RMB (Reator Multipropósito Brasileiro) é de, pelo menos, 80% em massa de U3Si2. No entanto, as técnicas de caracterização atualmente utilizadas no controle de qualidade não permitem quantificar as fases cristalinas diretamente. Assim, esse trabalho propõe a utilização da difração de raios X (DRX), alinhada a refinamento pelo método de Rietveld para caracterização do pó de siliceto. Para tal, foram produzidas ligas de urânio contendo 33 a 67 mol% de silício, e técnicas de moagem e ajustes de refinamento foram testados. O método desenvolvido inclui cominuição em moinho vibratório e DRX com refinamento automatizado dos dados, permitindo a quantificação das fases cristalinas de maneira confiável, rápida e com mínima interferência do operador. Os resultados obtidos foram corroborados com os de técnicas como análise de imagem obtida por microscópio eletrônica de varredura (MEV), densidade e análises elementares de U e Si. / The safe operation of a nuclear power system relies on the materials of its construction. During the lifetime of a nuclear power system, the materials are subject to high temperature, a corrosive environment, and damage from high-energy particles released during fission. The fuel which powers the reactor has a much shorter life, but is also subject to the harsh environments. Considering the several uranium alloys, the U3Si2 is largely used in research reactors, due to is high uranium density, high thermal conductivity, resistance to radiation-induced amorphization, swelling and crack propagation. During its fabrication by melting, however, more than one crystalline phase is usually formed, and, the behavior of each, under irradiation is different and possibly dangerous. For this reason, the specification of the IEA-R1 and RMB (Brazilian Multipurpose Reactor) nuclear reactors describes a minimum of 80wt.% of U3Si2 for the uranium silicide powder. In this sense, a quality control system is vital for the safety and performance of the reactor. Since the currently characterization techniques used do not quantify the crystalline phases directly, the present work proposes the use of X-ray diffraction (XRD), together with Rietveld refinement of the results, for uranium silicide powder characterization. To accomplish this objective, uranium allows were produced containing 33 to 67 mol% of silicon. Milling methods and refinements strategies were tested to improve XRD results. The proposed method includes vibration grinding and XRD with automatic refinement of results, producing fast, reliable and more unbiased results. The quantification results obtained were supported by other techniques as scanning electron microscopy image analysis, density and elementary U and Si characterization.

Made available in DSpace on 2014-10-09T12:34:19Z (GMT). No. of bitstreams: 0 / Made available in DSpace on 2014-10-09T14:10:09Z (GMT). No. of bitstreams: 0 / Dissertação (Mestrado) / IPEN/D / Instituto de Pesquisas Energeticas e Nucleares - IPEN-CNEN/SP

Made available in DSpace on 2014-10-09T12:34:19Z (GMT). No. of bitstreams: 0 / Made available in DSpace on 2014-10-09T14:10:09Z (GMT). No. of bitstreams: 0 / No início dos anos 90, os primeiros indícios do tráfico ilícito de materiais nucleares e radioativos foram observados principalmente na região européia. Uma década marcada por inúmeros casos de apreensão desse tipo de material. Como resultado, esses atos passaram a ser alvo de investigações criminais forenses, desenvolvendo-se a partir daí, a ciência forense nuclear. No Brasil não há registros oficiais do tráfico ilícito de material nuclear, entretanto, é amplamente conhecida a extração e o transporte ilegal de materiais geológicos radioativos, assim como a apreensão de fragmentos de materiais utilizados como blindagem de fontes radioativas. Uma das principais ferramentas utilizadas na ciência forense nuclear consiste no estabelecimento de bancos de dados de materiais nucleares. Esses documentos devem conter o maior número possível de informações sobre as características físicas, químicas e nucleares do material apreendido, permitindo a identificação de sua origem, processo de fabricação ou mesmo a idade (age). Assim, se estabelecem padrões de composição característicos de cada material, denominados assinaturas químicas (chemical finger print). Nesse trabalho foi adotado o protocolo forense nuclear seguindo as três etapas de avaliação sugeridas pela Agência Internacional de Energia Atômica (AIEA) na identificação da origem de Siliceto de urânio (U3Si2). Realizaram-se ensaios de caracterização física, química e isotópica dos materiais em estudo e compararam-se os dados com aqueles obtidos para outros compostos de urânio (Tetrafluoreto de urânio, UF4; óxidos de urânio, UO2 e U3O8; tricarbonato de amônio e uranila, TCAU) estabelecendo-se uma assinatura característica para cada um. A partir dos resultados foi possível classificar os compostos por grupos de origem, uma vez que são provenientes de diferentes processos de fabricação e/ ou origem. Demonstrou-se também a importância da criação e manutenção de um banco de dados nuclear na investigação de um evento forense nuclear. / Dissertação (Mestrado) / IPEN/D / Instituto de Pesquisas Energeticas e Nucleares - IPEN-CNEN/SP

During an extended LOCA in a LWR, the current UO2 fuel reaches very high temperatures and eventually melts, while the current Zircaloy fuel cladding oxidizes releasing hydrogen. These two consequences can lead to an unacceptable amount of radioactive release by presenting accident routes for containment failure. After such an accident at the Fukushima NPP in 2011, the development of Accident Tolerant Fuels LWRs gained additional momentum which aims to increase the margin to fuel melting, and to preserve cladding integrity as long as possible. Among the top ATF candidates compounds are UN and U3Si2, which have a high thermal conductivity and high uranium density. UN melts at 2850 °C on par with UO2, while U3Si2 melts at only 1665 °C. U3Si2 may serve as a second phase in UN–U3Si2 composites with better material properties than pure UN. Early studies on powders and dense samples, found the chemical UN corrosion by steam at all T,p pairs to generate a sandwiched UN/α‒U2N3+x/UO2 corrosion layer with inferior density. It was seen that dense polycrystalline UN would perform poorly due to an intergranular cracking mechanism due to the stresses caused by the growth of this layer. Due to the missing technological ability to control parameters like grain size and open porosity no work exist on the microstructure dependence of high density UN pellet corrosion in steam, and the intergranular cracking mechanism was never captured by imaging techniques. Also, as UN–U3Si2 composites are fairly new, the degradation mechanism of high density samples under steam is not known as no degradation study has yet been published. This work aimed at increasing understanding of the high pressure steam degradation mechanism of Spark Plasma sintered, microstructure controlled, UN and UN–10U3Si2 (wt %) composite samples, and to analyze the influence of grain size and density on the UN corrosion rate. A further goal was to image corrosion progress in the microstructure. For this, HPBAC steam exposure tests on UN and UN–10U3Si2 at 303 °C and 9 MPa were done. Durations were up to 1.5 hours. Samples were 96–99.9% of theoretical density and grain sizes were 6–24 µm. The corrosion in the microstructure of all samples is imaged by Light Optical Microscopy (LOM). Scanning Electron Microscopy/Energy Dispersive Spectroscopy (SEM/EDS) were used to track the change in chemical composition at the grain boundaries. Two continuous steam exposures in flowing Ar and N2 at 400 °C and 1 atm have been done to study the role of N2 and NH3 on the degradation. One TGA on the residue of one of the autoclave tests was done to confirm the final oxidation state. TGA confirms that at 303 °C and 9 MPa the final product is UO2, while Digerator results show that under N2 the corrosion is faster. LOM and SEM/EDS show that UN–pellets exposed to steam are breaking apart by intergranular cracks generated by a layered precipitation of U2N3+x/UO2 in the grain boundaries. As the density of the products differs greatly from that of UN, high intergranular stresses result in cracking. Cracking makes progressively more surfaces available to oxidation/hydrolysis. An increase in density and reduction of open porosity slows the corrosion process, while an increase in grain size accelerates the degradation. Consequently, all other considerations cast aside the most waterproofed microstructure of a pure polycrystalline UN sample will have maximized density, eliminated open porosity, while maintaining a small grain size. As clusters of UN grains are enveloped by the U3Si2 phase in UN–10%U3Si2, the cracking was seen to be predominantly intragranular. Irrespective of the quality of the microstructure polycrystalline UN will fail by intergranular corrosion. U3Si2 seems to react preferentially with the steam precipitating UO2, delaying the attack on the UN grains. The low degree of maximum weight gain and different corrosion progression in the microstructure of UN–10U3Si2 are strong indications that the composite may provide significantly higher steam tolerance than pure UN.

Theoretical models are used in support of the I2S-LWR (Integral Inherently Safe LWR) project for a direct comparison of fuel swelling and fission gas bubble formation between U₃Si₂ and UO₂ fuels. Uranium silicide is evaluated using a model developed by Dr. J. Rest with the fuel in a amorphous state. The uranium dioxide is examined with two separate models developed using a number of papers. One model calculates the swelling behavior with a fixed grain radius while the second incorporates grain growth into the model. Uranium silicide rapidly becomes amorphous under irradiation. The different mechanisms controlling the swelling of the fuels are introduced including the knee point caused by the amorphous state for the U₃Si₂. The outputs of each model are used to compare the fuels.

A demanda por combustível nuclear para reatores de pesquisa está aumentando em nível mundial, enquanto várias de suas fábricas têm pequeno volume de produção. Este trabalho estabeleceu um modelo conceitual com duas estratégias para o aumento da capacidade produtiva dessas fábricas. Foram abordadas as fábricas que produzem elementos combustíveis tipo placa carregados com LEU U3Si2-Al, tipicamente usados em reatores nucleares de pesquisa. A primeira estratégia baseia-se na literatura da área de administração da produção e é uma prática frequente nas fábricas em geral. A segunda estratégia aproveita a possibilidade de desmembrar setores produtivos, comum em instalações de produção de combustível nuclear. Ambas as estratégias geraram diferentes cenários de produção, os quais devem ser seguros em relação à criticalidade. Foram coletados dados de uma fábrica real de combustível nuclear para reatores de pesquisa. As duas estratégias foram aplicadas a esses dados com a finalidade de testar o modelo proposto, o que configurou um estudo de caso. A aplicação das estratégias aos dados coletados deu-se por meio de simulação de eventos discretos em computador. Foram criados diversos modelos de simulação para abranger todos os cenários gerados, de forma que o teste indicou um aumento da capacidade produtiva de até 207% sem necessidade de aquisição de novos equipamentos. Os resultados comprovam que o modelo atingiu plenamente o objetivo proposto. Como principal conclusão pode-se apontar a eficácia do modelo proposto, fato que foi validado pelos dados da fábrica. / Although many nuclear fuel factories have small production volumes, the demand for nuclear fuel for research reactors is increasing worldwide. This work established a conceptual model with two strategies to increase the production capacity of these factories. We addressed factories that produce plate-type fuel elements loaded with LEU U3Si2-Al, which are typically used in nuclear research reactors. The first strategy is based on production management literature and is a regular practice in general manufacturing plants. The second strategy takes advantage of the fact that productive sectors can be separated in nuclear fuel production facilities. Both strategies have generated different production scenarios that are assumed to be safe in relation to nuclear criticality. We collected data from a real plant that produces nuclear fuel for research reactors and applied the model to that data, aiming to test the proposed model by setting up a case study. Through the use of computer software, we applied the two strategies to this data by means of discrete events simulation and created several simulation models in order to cover all generated scenarios. Our tests indicated an increase of up to 207% in productive capacity without the need of acquiring new equipment, thus showing that the model has fully achieved its proposed objective. One of the main conclusions that we point out is the models effectiveness, which was validated by the factory data.

A demanda por combustível nuclear para reatores de pesquisa está aumentando em nível mundial, enquanto várias de suas fábricas têm pequeno volume de produção. Este trabalho estabeleceu um modelo conceitual com duas estratégias para o aumento da capacidade produtiva dessas fábricas. Foram abordadas as fábricas que produzem elementos combustíveis tipo placa carregados com LEU U3Si2-Al, tipicamente usados em reatores nucleares de pesquisa. A primeira estratégia baseia-se na literatura da área de administração da produção e é uma prática frequente nas fábricas em geral. A segunda estratégia aproveita a possibilidade de desmembrar setores produtivos, comum em instalações de produção de combustível nuclear. Ambas as estratégias geraram diferentes cenários de produção, os quais devem ser seguros em relação à criticalidade. Foram coletados dados de uma fábrica real de combustível nuclear para reatores de pesquisa. As duas estratégias foram aplicadas a esses dados com a finalidade de testar o modelo proposto, o que configurou um estudo de caso. A aplicação das estratégias aos dados coletados deu-se por meio de simulação de eventos discretos em computador. Foram criados diversos modelos de simulação para abranger todos os cenários gerados, de forma que o teste indicou um aumento da capacidade produtiva de até 207% sem necessidade de aquisição de novos equipamentos. Os resultados comprovam que o modelo atingiu plenamente o objetivo proposto. Como principal conclusão pode-se apontar a eficácia do modelo proposto, fato que foi validado pelos dados da fábrica. / Although many nuclear fuel factories have small production volumes, the demand for nuclear fuel for research reactors is increasing worldwide. This work established a conceptual model with two strategies to increase the production capacity of these factories. We addressed factories that produce plate-type fuel elements loaded with LEU U3Si2-Al, which are typically used in nuclear research reactors. The first strategy is based on production management literature and is a regular practice in general manufacturing plants. The second strategy takes advantage of the fact that productive sectors can be separated in nuclear fuel production facilities. Both strategies have generated different production scenarios that are assumed to be safe in relation to nuclear criticality. We collected data from a real plant that produces nuclear fuel for research reactors and applied the model to that data, aiming to test the proposed model by setting up a case study. Through the use of computer software, we applied the two strategies to this data by means of discrete events simulation and created several simulation models in order to cover all generated scenarios. Our tests indicated an increase of up to 207% in productive capacity without the need of acquiring new equipment, thus showing that the model has fully achieved its proposed objective. One of the main conclusions that we point out is the models effectiveness, which was validated by the factory data.