The nuclear fuel cycle is modelled as a cyclic, multi-stage production-to-inventory system. The objective is to meet a known deterministic demand for energy while minimizing acquisition, production, and inventory holding costs for all stages of the fuel cycle. The model allows for cyclic flow (feedback) of materials, material flow conversion factors at each stage, production lag times at each stage, and for escalating costs of uranium ore. It does not allow shortages to occur in inventories. The model is optimized by the application of the calculus of variations and specifically through recently developed theorems on the solution of functionals constrained by inequalities. The solution is a set of optimal cumulative production trajectories which define the stagewise production rates. Analysis of these production rates reveals the optimal nuclear fuel cycle costs and that inventories (stockpiles) occur in uranium fields, enriched uranium hexafluoride, and fabricated fuel assemblies. An analysis of the sensitivity of the model to variation in three important parameters is performed. / Ph. D.

Using the fuel burnup as objective function, it is desired to determine the initial distribution of the fuel in a reactor in order to obtain the maximum energy possible, for which, without changing a fixed initial fuel mass, the results for different initial fuel and control poison configurations are analyzed and the corresponding running times compared. One-dimensional, two energy-group theory is applied to a reflected cylindrical reactor using U-235 as fuel and light water as moderator and reflector. Fissions in both fast and thermal groups are considered. The reactor is divided into several annular regions, and the constant flux approximation in each depletion step is then used to solve the fuel and fission-product poisons differential equations in each region. The computer code OPTIME was developed to determine the time variation of core properties during the fuel cycle. At each depletion step, OPTIME calls ODMUG, [12] a criticality search program, from which the spatially-averaged neutron fluxes and control poison cross sections are obtained. A uniform initial fuel distribution was chosen as a benchmark and the results for several different fuel configurations were analyzed. Two different initial control poison distributions were investigated for each fuel configuration: a uniform and a fuel proportional distribution. Using an iterative approach fuel was moved from the low burnup regions toward the high burnup regions; reactor running times were in this way increased from 9000 to 11,500 hours in the fuel proportional control poison distribution case and from 9000 to 11,000 hours in the uniform control poison distribution case. Beyond this point not only did the running time not increase, but no criticality was reached. / Ph. D.

At a nuclear fuel reprocessing plant the special nuclear materials (SNM) are received as constituents of spent fuel assemblies, are converted to liquid form, and undergo a series of chemical processes. Uncertainties in measurements of SNM at each stage of reprocessing limit the accuracy of simple material balance accounting as a safeguards method. To be effective, a formal safeguards program must take into account all sources of measurement error yet detect any diversion of SNM. The objective of this study is to demonstrate an analytical method for assessing the accountability of selected constituent SNM. A combined discrete-continuous, time-dependent model using the GASP IV simulation language is developed to simulate mass flow, material accountability and measurement error at each stage of the reprocessing plant. The study demonstrates that the simulation method may be utilized to estimate the magnitude of SNM loss in an operating reprocessing plant which could reasonably be detected. Thus, the simulation method provides a level of confidence for effective loss/no-loss decisions. / Ph. D.

Nenhuma / Neste trabalho foi realizada a modelagem geométrica computacional das Cetapas de prensagem e sinterização da pastilha e da laminação da placa de combustível nuclear contendo microesferas de (Th,25%U)O2 dispersas em matriz de aço inoxidável com o objetivo de avaliar a distribuição destas microesferas nas diversas etapas do processamento. As regras de modelagem foram desenvolvidas baseadas nos parâmetros de cada etapa da fabricação da placa combustível. Para isto foram obtidas placas através do processamento por laminação de molduras de chapas de aço inoxidável, contendo pastilha fabricadas com microesferas de (Th,25%U)O2 com carregamentos de 10, 20 e 40% em peso de combustível disperso em matriz de aço inoxidável. Os dados das placas com carregamentos de 30 e 50% foram obtidos por interpolação da curva. As microesferas, obtidas pelo processo sol-gel, foram previamente secas, reduzidas e sinterizadas a 1700oC, durante 2 horas, sob atmosferas de hidrogênio. As microesferas sinterizadas alcançaram uma densidade de cerca de 98% da densidade teórica, e possuem um diâmetro médio de cerca de 300 mm e uma elevada resistência à fratura, de aproximadamente 40 N/microesfera. As regras implementadas neste modelo foram aplicadas nas coordenadas dos centros das esferas virtuais, que simulam as microesferas combustíveis de (Th,25%U)O2, obtendo-se novas coordenadas espaciais para cada uma delas nas etapas de prensagem e sinterização da pastilha e da laminação da placa combustível. Este modelo foi projetado com o uso de técnicas de análise de sistema estruturada, implementado utilizando a linguagem de programação Delphi e os resultados visualizados através do programa AutoCAD. Os resultados do modelo foram validados comparando-se as frações volumétricas experimentais em cada um dos carregamentos estudados com as frações simuladas. Este trabalho será de grande valia para o estudo do carregamento de microesferas na placa combustível, permitindo obter um combustível de elevado desempenho mecânico, térmico e neutrônico mesmo em mais alto carregamento. / The computational geometric modeling of the pressing, sintering and lamination stages for nuclear fuel plates composed by (Th,25%U)O2, microspheres dispersed into stainless steel matrix has been done in order to investigate the microspheres distribution in the various processing stages. The modeling standards were based on the parameters related to each fuel plate manufacturing stage. Accordingly, the plates were obtained through lamination processing of stainless steel plate frames comprising (Th,25%U)O2 microspheres pellets dispersed into stainless steel powder with loading of 10, 20 and 40% of microspheres dispersed into stainless steel matrix. The data for plates with loading of 30 and 50% have been obtained by linear interpolation. The microspheres produced by the sol-gel method were previously reduced and sintered at 1700 0C during 2 hours at hydrogen atmosphere. These sintered microspheres have reached about 98% of the theoretical density, with a mean diameter of 300 mm and a high resistance to fracture, near to 40 N/microsphere. The implemented standards in this model were applied at the virtual spheres center coordinates, which simulate the (Th,25%U)O2 fuel microspheres, and generate the new spatial coordinates to each of them in the pressing, sintering and lamination stages. This model was developed using structured system analysis techniques and it has been implemented using the Delphi programming language. The results were displayed through the AutoCAD program and validated comparing the experimental volumetric fractions in each of the studied loading, with the simulated fractions. The results indicate that this work could be a powerful tool in the investigation of microspheres loading in the fuel plate, allowing the attainment of a high mechanical and neutronic performance fuel, even for higher level loading.

Call number: LD2668 .T4 1985 M66 / Master of Science

Thesis (MScEng (Mechanical and Mechatronic Engineering))--University of Stellenbosch, 2010. / ENGLISH ABSTRACT: Regulators and designers of nuclear reactors regard knowledge of the pebble fuel temperature as important, due to the role that it plays in maintaining structural integrity and the production of neutrons. By using special fuel assemblies fitted with measuring equipment it is possible to measure the fuel temperature in stationary fuel reactors. This, however, is not possible in the pebble bed modular reactor due to its dynamic core. Designers of the pebble bed modular reactor have reserved special inspection channel borings inside the center reflector for fuel temperature measurement. By means of optical fibers and interferometry, the temperature can be measured inside such a channel. Currently the only way to control the fuel surface and core temperature is by measuring the gas inlet and outlet temperatures. This thesis attempts to determine the pebble temperature by measuring the temperature in a reflector channel. This is done by constructing an electrically heated pebble bed experimental setup simulating a cutout section of a pebble bed modular reactor core. An additional computational fluid dynamics simulation of the experimental setup was also performed. This thesis also attempts to determine if there is a measureable temperature peak that can indicate where a pebble was in contact with the reflector surface. This could then be used in future studies to determine the pebble fuel velocity as it moves down the reactor core. The computational fluid dynamics results were validated by experimental measurements. In the computational fluid dynamics model and experimental setup, it was found that there was indeed a measureable temperature difference on the temperature gradient along the reflector wall. The heat being conducted away from the pebble through the contact area can explain this. These differences were only observed when the channel was moved closer to the pebbles and it is therefore advised that some redesigning of the channel should be done if the in-core temperature is to be accurately interpreted by the designers at PBMR (Pty) Ltd. / AFRIKAANSE OPSOMMING: Reguleerders en ontwerpers van kern reaktore beskou die kennis van die korrel brandstof temperatuur as belangrik. Dit is weens die rol wat die brandstof temperatuur speel met die behoud van strukturele integriteit en die produksie van neutrone binne-in die reaktor. Met behulp van spesiale brandstof montasies toegerus met die meetings instrumentasie, is dit moontlik om die brandstof temperatuur in stilstaande brandstof reaktore te meet. Dit is egter nie moontlik in die korrel bed modulêre reaktor nie, as gevolg van sy dinamiese kern. Ontwerpers van die korrel bed modulêre reaktor het spesiale kanale in die binnekant van die middel reflektor vir brandstof temperatuur meeting gereseveer. Deur middel van optiese vesel en interferometrie, kan die temperatuur binne so 'n kanaal gemeet word. Tans is die enigste manier om die brandstof-oppervlak temperatuur te berekern, net moontlik deur gebruik te maak van die gemete gas inlaat-en uitlaat temperature van die reaktor. Hierdie tesis poog om vas te stel of die korrel brandstof temperatuur deur die meet van die oppervlak temperatuur in 'n reflektor-kanaal bepaal kan word. Dit word gedoen deur 'n elektriese verhitte korrel bed eksperimentele opstelling te bou wat 'n gedeelte van 'n korrel bed modulêre reaktor simuleer. 'n Bykomende numeriese simulasie van die eksperimentele opstelling was ook uitgevoer. Hierdie werk het ook probeer om vas te stel of daar 'n meetbare temperatuur piek op die temperatuur profiel aandui kan word waar 'n korrel in kontak is met die reflektor se oppervlak. Dit kan dan in toekomstige studies gebruik word om te bepaal wat die korrel brandstof spoed was soos dit in die reaktor beweeg. Die numerise simulasie uitslae was deur eksperimentele metings bevestig. In die numerise simulasie model en die eksperimentele opstelling, is daar gevind dat daar inderdaad 'n meetbare temperatuur verskil op die temperatuurgradiënt teen die reflektor oppervlak is. Dit kan verduidelik word as gevolg van die hitte wat weg van die korrel gelei word deur middel van die kontak area. Hierdie verskille was slegs waargeneem wanneer die kanaal nader aan die korrels geskuif is en dit word as n aanbeveling aan PBMR (Pty) Ltd gemaak om sommige herontwerpe aan die kanaal te doen indien die in-kerntemperatuur gemeet wil word en akkuraat geinterpreteer wil word.

Nenhuma / Uma das principais etapas na fabricação de combustíveis nucleares tipo placa, para a utilização em reatores de pesquisa e de propulsão naval, consiste no desenvolvimento de métodos e técnicas de ensaios não destrutivos para a avaliação do combustível nuclear durante a fabricação, assim como para análises do combustível pós-irradiação. Os ensaios não destrutivos podem contribuir para a detecção de descontinuidades durante as etapas de fabricação do combustível, como trincas e falhas na união entre o cerrne e o revestimento, que podem provocar a falha do combustível durante o seu uso em reatores nucleares. Métodos de ensaio como visual, radiográfico, correntes parasitas e ultra-som podem ser utilizados para essa finalidade. Neste trabalho foi abordado o uso do ensaio não destrutivo por ultra-som para a avaliação de combustíveis nucleares tipo placa. Devido às pequenas espessuras dos combustíveis tipo placa, assim como aos diferentes materiais presentes nos mesmos, foram utilizados, nos experimentos, transdutores ultra-sônicos de contato com sapatas de atraso e transdutores de imersão. Os ensaios foram realizados em um protótipo de combustível tipo placa constituído por um núcleo de UO2 disperso em uma matriz metálica de aço inoxidável, com revestimento em aço inoxidável. Neste protótipo foram usinados diferentes tipos de refletores artificiais, simulando a presença de descontinuidades naturais. Para os testes com os transdutores de imersão foi desenvolvido um dispositivo para a obtenção do perfil do feixe sônico emitido pelos mesmos, de forma a identificar a região de maior sensibilidade do feixe para o ensaio. Foram ainda fabricadas algumas lentes acústicas para a focalização do feixe, neste caso, sem sucesso. O uso dos diferentes tipos de transdutores ultra-sônicos possibilitou o estabelecimento de uma metodologia para a detecção de descontinuidades com diferentes geometrias e dimensões. O protótipo de combustível desenvolvido para os experimentos demonstrou ser adequado para estudos de sensibilidade do sistema de ensaio. / One of the most important steps in the fabrication processes of plate type nuclear fuels, intended to be used in research reactors or naval propulsion, is the development of nondestructive testing (NDT) methods and techniques for their quality assessment during fabrication and post-irradiation analysis. Those tests can contribute to detect discontinuities such as cracks and fails in meat-cladding junctions, that can lead to failures when installed and used in reactors. Examples of NDT methods that may be used for this purpose are visual inspection, radiography, eddy current and ultrasound. The objective of this study is to present the utilization of ultrasound methods to evaluate plate type nuclear fuels. Due to the small thicknesses of such kind of fuels, as well as the presence of different materials, the ultrasonic transducers used to perform the experiments were immersion type or contact with delay shims. Furthermore, a dummy plate fuel, constituted by a dispersion of UO2 in stainless steel matrix, with stainless steel cladding, was specially constructed. In the surface of such plate, several kinds of artificial reflectors, simulating the presence of natural flaws were machined. For immersion type ultrasonic transducers, a mechanical scanning system was developed to allow the determination of their sonic beam profiles and identification of the highest sensitivity beam region. Additionally, some acoustic lenses, useful to help on beam focalization, were fabricated and used, but the expected performance was not achieved. The use of different kinds of ultrasonic transducers allowed the establishing of a methodology to detect discontinuities of different geometry and sizes. The developed dummy fuel demonstrated to be adequate for the studies of sensitivity of the test system.

Coordenação de Aperfeiçoamento de Pessoal de Nível Superior / Neste trabalho foram realizadas a cominuição e passivação da liga metálica U-4Zr-2Nb pelo processo de hidretação-desidretação, bem como o estudo da sua cinética. A obtenção deste material pulverizado através das técnicas da metalurgia do pó é uma etapa necessária e chave na fabricação da pastilha, que será empregada na laminação da placa combustível. Foi escolhida a liga com composição U-4Zr-2Nb devido à sua elevada densidade e baixo teor de elementos de liga, além de suas pequenas seções de choque para nêutrons térmicos. Previamente, foi projetado e construído o equipamento tipo Sievert volumétrico para a cominuição da liga metálica de urânio pelo processo de hidretação-desidretação, operacionalizando-o no modo automático, através da aquisição de dados por intermédio de softwares também desenvolvidos neste trabalho. Juntamente com o desenvolvimento deste equipamento, outro software foi desenvolvido para calcular a cinética de hidretração e a porcentagem hidretada. A seguir, com a utilização deste equipamento, amostras da liga U-4Zr-2Nb foram tratadas termicamente, hidretadas, passivadas, moídas e desidretadas. O processo de cominuição desenvolvido foi realizado nas condições de temperaturas de hidretação variando entre 108C e 295C e a pressão variando entre 2,0 bar e 1,5 bar. Todas as amostras foram hidretadas por completo, independentemente da temperatura de processamento. O tempo de hidretação variou entre 550 a 16176 segundos, de acordo com a temperatura utilizada, sendo mais rápido para temperaturas mais altas. Independentemente dos tratamentos térmicos feitos previamente nas amostras, todas apresentaram somente a fase α e, conseqüentemente, todas as hidretações realizadas foram massivas. Foi desenvolvido também um processo de passivação dos pós obtidos, tendo-se conseguido amostras cominuídas estáveis, ou seja, não apresentaram reações pirofóricas quando expostas ao ar, nem uma excessiva oxidação das mesmas. Para isto, foi utilizada uma mistura de gases contendo 90% de argônio e 10% de oxigênio. Após a passivação, os hidretos foram moídos e passivados novamente para obtenção final do pó metálico. A granulometria final dos pós metálicos obtidos não depende dos tratamentos térmicos da amostra nem da temperatura de hidretação. As partículas maiores se revelaram um aglomerado de partículas menores e, portanto, foi utilizado um processo de moagem para desaglomeração parcial destas partículas, tendo-se obtidos partículas com tamanhos na faixa entre 11,2 e 22,4 μm. / In this work the comminution and passivation of U-4Zr-2Nb alloys by hydrading-dehydrading process was carried out and the kinetics of hydride formation was studied. The obtaining of the powdered material through the techniques of powder metallurgy is a key and necessary step in the manufacture of the pellet useful for providing the fabrication of the fuel plate. An alloy with composition U-4Zr-2Nb was chosen due to their high density and low alloying elements, in addition to its low thermal neutrons cross section. A volumetric Sievert equipment for comminuition of uranium alloys by the process of hydriding-dehydriding was designed and constructed. This equipment operates in an automatic mode through the data acquisition software also developed in this work. Along with the development of this equipment, other software was developed to calculate the kinetic of hydriding and the hydriding amount. Then, using this equipment, samples of the U-4Zr-2Nb alloy were heat treated, hydrided, passivated, milled and dehydrided. The developed comminution process was obtained in the temperature range of 108oC to 295oC and in the pressure range of 1.5 Bar to 2 Bar. All samples were completely hydrided, regardless of the hydriding temperature. The hydriding time ranged from 540 to 16176 seconds, according to the temperature used, being faster at higher temperature. Regardless of the previously heat treatments, all samples showed only the α phase and, consequently, all hydridings were massive performed. It was also developed a passivation process of the obtained powder, and the powdered samples were stable, not pyrophoric and no kind of reaction was observed when exposed to air, without an excessive oxidation. In this case, it was used a gas mixture of 90% argon and 10% oxygen. After passivation, the hydride were milled and passivated again to obtain the metallic powder. The final size of the powdered metal did not depend on the heat treatment of the sample or on the hydriding temperature. The larger particles revealed to be an agglomerate of particles and therefore the milling process partially dismantle these agglomerates into primary particles. The particles size ranged from 11.2 up to 22.4 μm.

A new generation of compact and highly efficient power production and propulsion technologies are critically needed in enabling NASAs long-term goals. Nuclear fission power technologies as part of project Prometheus are in development to meet this need. Proposed reactor concepts utilize a combination of refractory metals and stainless steels. One such refractory alloy, Niobium 1% Zirconium (Nb-1Zr), will be used because of its strength at high temperatures, neutron absorption properties, and resistance to corrosion by liquid alkali metals. One potential problem in using Nb-1Zr is that it undergoes rapid high temperature oxidation, even in low oxygen concentrations. Long-term oxidation of the niobium matrix can significantly deteriorate the mechanical properties of the alloy. This thesis reports on experimental studies of the high temperature interaction of 316 stainless steel (316 SS) and Nb-1Zr under prototypic space fission reactor operating conditions. Specifically, how the high temperature oxidation rate of Nb-1Zr changes when in contact with 316 SS at low external oxygen concentrations. The objective of the project is to determine if transport of gaseous contaminants, such as oxygen, will occur when Nb-1Zr is in contact with 316 SS, thereby increasing the oxidation rate and degrading material properties. Experiments were preformed in a realistic non-nuclear environment at the appropriate operating conditions. Thermal Gravimetric Analysis techniques were used to quantify results. Coupons of Nb-1Zr and Nb-1Zr in contact with 316 SS foil are subjected to flowing argon with oxygen concentrations between 4-15ppm and heated to a temperature of 500, 750, and 1000oC for 2 to 10 hours. Experiments were conducted at the Early Flight Fission Test Facility at NASA Marshall Space Flight Center. The experimental results indicate that a complex oxidation process, which depends greatly on temperature and oxygen concentration, occurs at the expected operating conditions. Non-linear regression techniques were applied to experimental data in order to derive correlations for the approximate oxidation rate of Nb-1Zr and Nb-1Zr in contact with 316 SS as a function of time, temperature, and oxygen concentration.

Recent nuclear proliferation concerns and disarmament agreements have encouraged the U.S. to decrease the excess amount of weapons-grade and reactor-grade plutonium. Continued use of nuclear power without a permanent solution for waste disposition has also led to the need for a reliable method by which the waste products, specifically plutonium, can be utilized or destroyed. One possible solution to plutonium destruction is achieved by manufacturing it into small microspheres and embedding it within an inert metal matrix, then placing it inside a conventional nuclear reactor. This process would burn some of the plutonium while producing electricity. PuO2Zr dispersion fuel has been proposed for such a purpose. Prior to its use, however, this non-fertile metal matrix dispersion fuel must be shown to be mechanically stable in the reactor environment. The internal mechanical interactions of dispersion fuel were modeled using finite element analysis. The results were used to assess the stability of PuO2Zr dispersion fuel inside a reactor. Several parameters, including fuel particle size, volumetric loading, temperature, and burnup, were varied to determine the maximum amount of plutonium that can be burned while maintaining fuel integrity. Earlier experiments using UO2 stainless steel dispersion fuels were used to validate the model and establish a failure criterion. The validated model was then used to determine the parameter space over which PuO2Zr dispersion fuel can be successfully used. These results show that PuO2Zr dispersion fuel is robust and may offer a reliable method for plutonium disposal in current reactors.