Optimization methods have been developed and refined throughout many scientific fields of study. This work utilizes one such developed technique of optimization called simulated annealing to produce optimal operation parameters for a 15x15 fuel assembly to be used in an operating nuclear power reactor. The two main cases of optimization are: one that finds the optimal 235U enrichment layout of the fuel pins in the assembly and another that finds both the optimal 235U enrichments where gadolinium burnable absorber pins are also inserted. Both of these optimizations can be performed by coupling Adaptive Simulated Annealing to TransLAT which successfully searches the optimization space for a fuel assembly layout that produces the minimized pin power peaking factor. Within given time constraints this package produces optimal layouts within a given set of assumptions and constraints. Each layout is forced to maintain the fuel assembly average 235U enrichment as a constraint. Reductions in peaking factors that are produced through this method are on the order of 2% to 3% when compared to the baseline results. As with any simulated annealing approach, families of optimal layouts are produced that can be used at the engineer’s discretion.

The nuclear waste currently generated in the United States is stored in spent fuel pools and dry casks throughout the country awaiting a permanent disposal solution. One efficient solution would be to remove the actinides from the waste and transmute these isotopes in a fast spectrum reactor. Currently this technology is unavailable on a commercial scale and a considerable amount of research and development is still required. An alternate solution is to reprocess and recycle the used fuel in thermal reactors, creating new fuel while reducing the amount of waste and its impact to the environment. This thesis examines the possibility of multi-recycling the transuranics (Pu, Np, Am, and Cm) in a standard pressurized water reactor (PWR). Two types of recycling strategies will be examined: one where Pu, Np, and Am are recycled (TRU-Cm) and a second where the previous isotopes as well as Cm are recycled (TRU+Cm). To offset the hardened neutron spectrum that results from the inclusion of the transuranics, a smaller fuel pin is employed to provide additional moderation. Computer simulations are used to model the in-reactor physics and long-term isotopic decay. Each fuel type is assessed based on the required U-235 enrichment, void coefficient, transuranic production/destruction, and radiotoxicity reduction as compared to a UOX and MOX assembly. It is found that the most beneficial recycling strategy is the one where all of the transuranics are recycled. The inclusion of Cm reduces the required U-235 enrichment, compared to the other multi-recycled fuel and, after a significant number of recycles, can result in the required enrichment to decrease. This fuel type also maintains a negative void coefficient for each recycle. The void coefficient of the fuel type without Cm becomes positive after the third cycle. The transmutation destruction of the two multi-recycled assemblies is less than that of a MOX assembly, but the transmutation efficiency of the multi-recycled assemblies exceeds the MOX assemblies. The radiotoxicity of both multi-recycled assemblies is significantly lower than the UOX and MOX with the TRU+Cm fuel being the lowest. When Curium is recycled only 28,000 years are required for the radiotoxicity of the waste to reach that of natural Uranium and when Cm is not recycled, the amount of time increases to 57,000 years.

Após o aumento de potência do reator IEA-R1 de 2 MW para 5 MW observou-se um aumento da taxa de corrosão nas placas laterais de alguns elementos combustíveis e algumas dúvidas surgiram com relação ao valor de vazão utilizada nas análises termo-hidráulicas. A fim de esclarecer e medir a distribuição de vazão real pelos elementos combustíveis que compõe o núcleo do reator IEA-R1, um elemento combustível protótipo, sem material nuclear, chamado DMPV-01 (Dispositivo para Medida de Pressão e Vazão), em escala real, foi projetado e construído em alumínio. A vazão no canal entre dois elementos combustíveis é muito difícil de estimar ou ser medida. Esta vazão é muito importante no processo de resfriamento das placas laterais. Este trabalho apresenta a concepção e construção de um elemento combustível instrumentado para medir a temperatura real nestas placas laterais para melhor avaliar as condições de resfriamento do combustível. Quatorze termopares foram instalados neste elemento combustível instrumentado. Quatro termopares em cada canal lateral e quatro no canal central, além de um termopar no bocal de entrada e outro no bocal de saída do elemento. Existem três termopares para medida de temperatura do revestimento e um para a temperatura do fluido em cada canal. Três séries de experimentos, para três configurações distintas, foram realizadas com o elemento combustível instrumentado. Em dois experimentos uma caixa de alumínio foi instalada ao redor do núcleo para reduzir o escoamento transverso entre os elementos combustíveis e medir o impacto na temperatura das placas externas. Dada a tamanha quantidade de informações obtidas e sua utilidade no projeto, melhoria e capacitação na construção, montagem e fabricação de elementos combustíveis instrumentados, este projeto constitui um importante marco no estudo de núcleos de reatores de pesquisa. As soluções propostas podem ser amplamente utilizadas para outros reatores de pesquisa. / After the IEA-R1 upgrade from 2 MW to 5 MW it was observed that the corrosion rate increased in a lateral plate of one fuel element and some issues appeared concerning the flow values used in the thermal-hydraulic analysis. In order to clear it up and measure the actual flow distribution among the fuel elements composing the IEA-R1 active core, a dummy element without nuclear fuel material, called DMPV-01 (Pressure and Flow Measurement Device), full scale, was designed and manufactured in aluminum. The flow rate in the channel between two fuel assemblies is very difficult to estimate or measure. This flow rate is very important to the cooling process of the external plates. This work presents the design and construction of an instrumented fuel assembly in order to measure the actual temperature in these lateral plates. Fourteen thermocouples were installed in this instrumented fuel assembly. Four in each lateral channel, one in the inlet nozzle and one in the outlet nozzle. There are three thermocouples in each channel to measure the clad temperature and one thermocouple to measure the fluid temperature. Three series of experiments, for three different core configuration were carried out with the instrumented fuel assembly. In two experiments a box was installed around the core to reduce the cross flow between the fuel assembly and measure the impact in the temperatures of external plates. Given the amount of information generated and its utility in the design, improvement and qualification in construction, assembly and manufacturing of instrumented fuel, this project turned out to be an important landmark on the thermal-hydraulic study of research reactor cores. The proposed solutions could be useful for other research reactors.

Após o aumento de potência do reator IEA-R1 de 2 MW para 5 MW observou-se um aumento da taxa de corrosão nas placas laterais de alguns elementos combustíveis e algumas dúvidas surgiram com relação ao valor de vazão utilizada nas análises termo-hidráulicas. A fim de esclarecer e medir a distribuição de vazão real pelos elementos combustíveis que compõe o núcleo do reator IEA-R1, um elemento combustível protótipo, sem material nuclear, chamado DMPV-01 (Dispositivo para Medida de Pressão e Vazão), em escala real, foi projetado e construído em alumínio. A vazão no canal entre dois elementos combustíveis é muito difícil de estimar ou ser medida. Esta vazão é muito importante no processo de resfriamento das placas laterais. Este trabalho apresenta a concepção e construção de um elemento combustível instrumentado para medir a temperatura real nestas placas laterais para melhor avaliar as condições de resfriamento do combustível. Quatorze termopares foram instalados neste elemento combustível instrumentado. Quatro termopares em cada canal lateral e quatro no canal central, além de um termopar no bocal de entrada e outro no bocal de saída do elemento. Existem três termopares para medida de temperatura do revestimento e um para a temperatura do fluido em cada canal. Três séries de experimentos, para três configurações distintas, foram realizadas com o elemento combustível instrumentado. Em dois experimentos uma caixa de alumínio foi instalada ao redor do núcleo para reduzir o escoamento transverso entre os elementos combustíveis e medir o impacto na temperatura das placas externas. Dada a tamanha quantidade de informações obtidas e sua utilidade no projeto, melhoria e capacitação na construção, montagem e fabricação de elementos combustíveis instrumentados, este projeto constitui um importante marco no estudo de núcleos de reatores de pesquisa. As soluções propostas podem ser amplamente utilizadas para outros reatores de pesquisa. / After the IEA-R1 upgrade from 2 MW to 5 MW it was observed that the corrosion rate increased in a lateral plate of one fuel element and some issues appeared concerning the flow values used in the thermal-hydraulic analysis. In order to clear it up and measure the actual flow distribution among the fuel elements composing the IEA-R1 active core, a dummy element without nuclear fuel material, called DMPV-01 (Pressure and Flow Measurement Device), full scale, was designed and manufactured in aluminum. The flow rate in the channel between two fuel assemblies is very difficult to estimate or measure. This flow rate is very important to the cooling process of the external plates. This work presents the design and construction of an instrumented fuel assembly in order to measure the actual temperature in these lateral plates. Fourteen thermocouples were installed in this instrumented fuel assembly. Four in each lateral channel, one in the inlet nozzle and one in the outlet nozzle. There are three thermocouples in each channel to measure the clad temperature and one thermocouple to measure the fluid temperature. Three series of experiments, for three different core configuration were carried out with the instrumented fuel assembly. In two experiments a box was installed around the core to reduce the cross flow between the fuel assembly and measure the impact in the temperatures of external plates. Given the amount of information generated and its utility in the design, improvement and qualification in construction, assembly and manufacturing of instrumented fuel, this project turned out to be an important landmark on the thermal-hydraulic study of research reactor cores. The proposed solutions could be useful for other research reactors.

Deformations of nuclear fuel assemblies have been observed in nuclear power plants since the mid-90s. Such deformations are generally called bowing effects. Fuel assemblies under high irradiation undergo growth and creep induced by high loading forces and low skeleton stiffness of the assemblies which gives permanent deformations and modifies moderation regions. Hence, giving an unpredicted neutron flux spectrum, power distribution, and isotopic concentrations in the burnt fuel. The aim of this thesis is to study the effects of local fuel bowing in terms of power distribution and isotopic composition changes through simulations of the reactor core.  The reactor is simulated with realistic bowing maps and previous deterministically simulated realistic fuel data from a present reactor by deploying the Monte Carlo method using the nuclear reactor code Serpent 2. Two subparts of a full reactor core with fuel from separate fuel cycles are investigated in 2D using burnup. To quantify the impact of the bowing, the change in power distribution and the induced isotopic composition change are calculated by a relative difference between a nominal case and a simulation with perturbed fuel assemblies. The results are presented in colormaps, for visualization. The isotopic composition for U235, U238, Pu239, Nd148, and Cm244 are investigated. Also, statistical uncertainty estimations in the composition of the depleted fuel are done by multiple calculations of the same geometry while changing the seed of random variables in the Monte Carlo calculation. The mean value and the standard deviation in the mass density of U235 and Pu239 are calculated for two pins together with histograms with a normal fit for each case to clarify the mathematical distribution of the calculations.  The simulations performed in this thesis have detected clear impacts of the reactor behavior in terms of power distribution and isotopic composition in the burnt fuel introduced by the bowing. Assembly perturbations of about 10 mm may locally introduce a 10 % relative difference in power density and U235 content between the nominal and the bowed case at 15 MWd/kgU burnup. The power and the isotopic composition changes agree with expectations from the bowing maps.

One of the greatest challenges facing the nuclear power industry is the final disposition of nuclear waste. To meet the needs of the nuclear power industry, a new fuel assembly design, called DUPLEX, has been developed which provides higher fuel burnups, burns transuranic waste while reducing minor actinides, reduces the long term radiotoxicity of spent nuclear fuel, and was developed for use in current light water reactors. The DUPLEX design considered in this thesis is based on a seed and blanket unit (SBU) configuration, where the seed region contains standard UO2 fuel, and the blanket region contains an inert matrix (Pu,Np,Am)O2-MgO-ZrO2 fuel. The research efforts of this thesis are first to consider the higher burnup effects on DUPLEX assembly thermal-hydraulic performance and thermal safety margin over the assembly’s expected operational lifetime. In order to accomplish this, an existing burnup-dependent thermal-hydraulic methodology for conventional homogeneous fuel assemblies has been updated to meet the modeling needs specific to SBU-type assemblies. The developed framework dramatically expands the capabilities of the latest thermal-hydraulic evaluation framework such that the most promising and unique DUPLEX fuel design can be evaluated. As part of this updated methodology, the posed DUPLEX design is evaluated with respect to the minimum departure from nucleate boiling ratio, peak fuel temperatures for both regions, and the peak cladding temperatures, under ANS Condition I, II, and III transient events with the thermal-hydraulic code VIPRE-01. Due to difficulty in the fabrication and handling of minor actinide dioxides, documented thermal conductivity values for the considered IMF design are unavailable. In order to develop a representative thermal conductivity model for use in VIPRE-01, an extensive literature survey on the thermal conductivity of (Pu,Np,Am)O2-MgO-ZrO2 component materials and a comprehensive review of combinatory models was performed. Using the updated methodology, VIPRE-01 is used to perform steady-state and transient thermal hydraulic analyses for the DUPLEX fuel assembly. During loss-of-flow accident scenarios, the DUPLEX design is shown to meet imposed safety criteria. However, using the most conservative thermal conductivity modeling approach for (Pu,Np,Am)O2-MgO-ZrO2, the blanket region fuel temperatures remain only slightly below the design limit.

Two new fuel assembly designs for light water reactors using advanced mixed-oxide fuels have been proposed to reduce the radiotoxicity of used nuclear fuel discharged from nuclear power plants. The research efforts of this thesis are the first to consider the effects of burnup on advanced mixed-oxide fuel assembly performance and thermal safety margin over an assembly?s expected operational burnup lifetime. In order to accomplish this, a new burnup-dependent thermal-hydraulic analysis methodology has been developed. The new methodology models many of the effects of burnup on an assembly design by including burnup-dependent variations in fuel pin relative power from neutronic calculations, assembly power reductions due to fissile content depletion and core reshuffling, and fuel material thermal-physical properties. Additionally, a text-based coupling method is developed to facilitate the exchange of information between the neutronic code DRAGON and thermal-hydraulic code VIPRE-01. The new methodology effectively covers the entire assembly burnup lifetime and evaluates the thermal-hydraulic performance against ANS Condition I, II, and III events with respect to the minimum departure from nucleate boiling ratio, peak cladding temperatures, and fuel centerline temperatures. A comprehensive literature survey on the thermal conductivity of posed fuel materials with burnup-dependence has been carried out to model the advanced materials in the thermal-hydraulic code VIPRE-01. Where documented conductivity values are not available, a simplified method for estimating the thermal conductivity has been developed. The new thermal conductivity models are based on established FRAPCON-3 fuel property models used in the nuclear industry, with small adjustments having been made to account for actinide additions. Steady-state and transient thermal-hydraulic analyses are performed with VIPRE- 01 for a reference UO2 assembly design, and two advanced mixed-oxide fuel assembly designs using the new burnup-dependent thermal-hydraulic analysis methodology. All three designs maintain a sufficiently large thermal margin with respect to the minimum departure from nucleate boiling ratio, and maximum cladding and fuel temperatures during partial and complete loss-of-flow accident scenarios. The presence of a thin (Am,Zr)O2 outer layer on the fuel pellet in the two advanced mixed-oxide fuel assembly designs increases maximum fuel temperatures during transient conditions, but does not otherwise greatly compromise the thermal margin of the new designs.

Les phénomènes d’interaction fluide-structure jouent un rôle important dans le calcul de tenue au séisme des assemblages combustibles. Afin de quantifier les marges de dimensionnement, le modèle de forces fluides utilisé doit être validé et affiné. Pour cela, des campagnes d’essais à l’échelle industrielle ont été réalisées en amont de la thèse. L’objectif ici est de contribuer à l’interprétation des essais industriels pour le cas stationnaire, et de valider les méthodes numériques permettant de simuler ce type d’écoulement. La problématique industrielle s'inscrit dans la tradition de l'étude des structures élancées sous écoulement axial. Le modèle de force fluide locale généralement utilisé, que nous appelons modèle de Taylor-Lighthill-Païdoussis (TLP), consiste en stationnaire à combiner un terme de force fluide potentielle, proportionnel à la courbure, et un terme de force fluide visqueuse, proportionnel à la pente. Des versions dynamiques de ce modèle ont été employées avec succès pour prédire le comportement vibratoire de cylindres flexibles en écoulement axial. Néanmoins, la littérature propose très peu de données de validation directe de cette représentation des forces fluides. Afin d’acquérir de telles données, pour le cas particulier d’un cylindre confiné dans un réseau de cylindres, un nouveau banc d’essai a été conçu et mis en place au laboratoire. Il s’agit d’un faisceau de 3x3 cylindres disposé dans une veine de soufflerie. Le cylindre central possède trois degrés de liberté : rotation, translation, flexion. Les efforts fluides résultants sont mesurés à l’aide d’une balance. Un modèle numérique similaire à la maquette est aussi réalisé et donne accès aux forces fluides locales. Les forces globales obtenues numériquement et expérimentalement sont comparables. Les forces locales obtenues dans les simulations numériques s’expliquent bien à l’aide du modèle TLP, en ignorant les effets de bord à l’entrée et à la sortie du faisceau. La transposition au cas industriel, de géométrie plus complexe, est réalisable par recalage des coefficients du modèle. / Fluid-structure interaction phenomena play a major role in the seismic design of fuel assemblies. In order to evaluate the design margins, the implemented model of fluid forces needs to be carefully assessed. Industrial-scale tests have been carried out with that purpose. Our goal is to contribute to their interpretation in the steady case, and to validate CFD methods usually applied to the type of flow at stake here. This fits in the tradition of the study of slender structures in axial flow. The local steady fluid forces decompose in a potential term, which is proportional to the curvature of the structure, and a viscous term, proportional to the angle of incidence. Adapted versions of this representation, which we call Taylor-Lighthill-Païdoussis (TLP) model, have proved successful in predicting the dynamic behaviour of flexible cylinders in axial flow. However, there is a lack in the literature of sound validation data for the fluid forces themselves. In order to gather such data, a new test rig has been designed and built. It consists in a 3x3 cylinder bundle confined in a wind tunnel. The central cylinder can be rotated, translated or bent. Resultant fluid forces are measured using a load cell. CFD calculations give access to the local fluid forces. CFD and experiments give similar results on the global fluid forces. The TLP model performs well at predicting the local fluid forces, except in the inlet and outlet regions. It can be fitted to the industrial case by adapting its coefficients.

The accumulation of spent nuclear fuel may be hindering the expansion of nuclear electricity production. However, the reprocessing and recycling of spent fuel may reduce its volume and environmental burden. Although fast spectrum reactors are the preferred modality for transuranic element transmutation, such fast spectrum systems are in very short supply. It is therefore legitimate to investigate the recycling potential of thermal spectrum systems, which constitute the overwhelming majority of nuclear power plants worldwide. To do so efficiently, several new fuel assembly designs are proposed in this Thesis: these include (1) Mixed Oxide Fuel (MOX), (2) MOX fuel with Americium coating, (3) Inert-Matrix Fuel (IMF) with UOX as inner zone, and (4) IMF with MOX as inner zone. All these designs are investigated in a multi-recycling strategy, whereby the spent fuel from a given generation is re-used for the next generation. The accumulation of spent nuclear fuel may be hindering the expansion of nuclear electricity production. However, the reprocessing and recycling of spent fuel may reduce its volume and environmental burden. Although fast spectrum reactors are the preferred modality for transuranic element transmutation, such fast spectrum systems are in very short supply. It is therefore legitimate to investigate the recycling potential of thermal spectrum systems, which constitute the overwhelming majority of nuclear power plants worldwide. To do so efficiently, several new fuel assembly designs are proposed in this Thesis: these include (1) Mixed Oxide Fuel (MOX), (2) MOX fuel with Americium coating, (3) Inert-Matrix Fuel (IMF) with UOX as inner zone, and (4) IMF with MOX as inner zone. All these designs are investigated in a multi-recycling strategy, whereby the spent fuel from a given generation is re-used for the next generation.

Experiments investigating complex flows in rod bundles with spacer grids that have mixing devices (such as flow mixing vanes) have mostly been performed using single-point measurements. Although these measurements allow local comparisons of experimental and numerical data they provide little insight because the discrepancies can be due to the integrated effects of many complex flow phenomena such as wake-wake, wake-vane, and vane-boundary layer interactions occurring simultaneously in a complex flow environment. In order to validate the simulations results, detailed comparison with experimental data must be done. This work describes an experimental database obtained using Time Resolved Particle Image Velocimetry (TR-PIV) measurements within a 5 x 5 rod bundle with spacer-grids. Measurements were performed using two different grid designs. One typical of Boiling Water Reactors (BWR) with swirl type mixing vanes and the other typical of Pressurized Water Reactors (PWR) with split type mixing vanes. High quality data was obtained in the vicinity of the grid using the multi-scale approach. One of the unique characteristic of this set-up is the use of the Matched Index of Refraction (MIR) technique employed in this investigation. This approach allows the use of high temporal and spatial non-intrusive dynamic measurement techniques to investigate the flow evolution below and immediately above the spacer. The experimental data presented includes explanation of the various cases tested such as test rig dimensions, measurement zones, the test equipment and the boundary conditions in order to provide appropriate data for comparison with Computational Fluid Dynamics (CFD) simulations. Turbulence parameters of the obtained data are analyzed in order to gain insight of the physical phenomena. The shape of the velocity profile at various distances from the spacer show important modifications passing the grid which delineates the significant effects of the presence of the grid spacer. Influence of the vanes wake in the global velocity was quantified to be up to a distance of 4 hydraulic diameters from the edge of the grid.Spatial and temporal correlations in the two measured dimensions were performed to quantify the time and length scales present in the flow in the vicinity of the grids and its influence in the flow modification induced by the vanes. Detection of vortex cores was performed using the vorticity, swirl strength and Galilean decomposition approach. The resulted cores were then tracked in time, in order to observe the evolution of the structures under the influence of the vanes for each grid. Vortex stretching was quantified in order to gain insight of the energy dissipation process normally associated with the phenomena. This work presents data in a single-phase flow situation and an analysis of these data for understanding complex flow structure. This data provide for the first time detailed temporal velocity full field which can be used to validate CFD codes.