In this thesis a numerical study has been undertaken to investigate turbulent flow and heat transfer in a number of flow problems, representing the gas-cooled reactor core flows. The first part of the research consisted of a meticulous assessment of various advanced RANS models of fluid turbulence against experimental and numerical data for buoyancy-modified mixed convection flows, such flows being representative of low-flow-rate flows in the cores of nuclear reactors, both presently-operating Advanced Gas-cooled Reactors (AGRs) and proposed ‘Generation IV’ designs. For this part of the project, an in-house code (‘CONVERT’), a commercial CFD package (‘STAR-CD’) and an industrial code (‘Code_Saturne’) were used to generate results. Wide variations in turbulence model performance were identified. Comparison with the DNS data showed that the Launder-Sharma model best captures the phenomenon of heat transfer impairment that occurs in the ascending flow case; v^2-f formulations also performed well. The k-omega-SST model was found to be in the poorest agreement with the data. Cross-code comparison was also carried out and satisfactory agreement was found between the results.The research described above concerned flow in smooth passages; a second distinct contribution made in this thesis concerned the thermal-hydraulic performance of rib-roughened surfaces, these being representative of the fuel elements employed in the UK fleet of AGRs. All computations in this part of the study were undertaken using STAR-CD. This part of the research took four continuous and four discrete design factors into consideration including the effects of rib profile, rib height-to-channel height ratio, rib width-to-height ratio, rib pitch-to-height ratio, and Reynolds number. For each design factor, the optimum configuration was identified using the ‘efficiency index’. Through comparison with experimental data, the performance of different RANS turbulence models was also assessed. Of the four models, the v^2-f was found to be in the best agreement with the experimental data as, to a somewhat lesser degree were the results of the k-omega-SST model. The k-epsilon and Suga models, however, performed poorly. Structured and unstructured meshes were also compared, where some discrepancies were found, especially in the heat transfer results. The final stage of the study involved a simulation of a simplified 3-dimensional representation of an AGR fuel element using a 30 degree sector configuration. The v^2-f model was employed and comparison was made against the results of a 2D rib-roughened channel in order to assess the validity and relevance of the precursor 2D simulations of rib-roughened channels. It was shown that although a 2D approach is extremely useful and economical for ‘parametric studies’, it does not provide an accurate representation of a 3D fuel element configuration, especially for the velocity and pressure coefficient distributions, where large discrepancies were found between the results of the 2D channel and azimuthal planes of the 3D configuration.

No description available.

No description available.

Tecnécio-99m (99mTc), o produto de decaimento do molibdênio-99 (99Mo), é um dos radioisótopos mais utilizados na medicina nuclear, abrangendo cerca de 80% de todos os procedimentos de radiodiagnóstico médico pelo mundo. Atualmente o Brasil necessita de uma quantidade de aproximadamente 450 Ci de 99Mo por semana. Devido à crise e à escassez em seu fornecimento que vem sendo observada no cenário mundial desde 2008, o IPEN decidiu desenvolver um projeto próprio para produção de 99Mo através da fissão do urânio-235. O objetivo deste trabalho de dissertação foi desenvolver cálculos neutrônicos e temo-hidráulicos para avaliar a segurança operacional de um dispositivo para produção de 99Mo a ser irradiado no núcleo do reator IEA-R1. Neste dispositivo serão alojados dez alvos do tipo dispersão de UAlx-Al com baixo enriquecimento de urânio (LEU) e densidade de 2,889 gU/cm³. Para o cálculo neutrônico foram utilizados os programas computacionais HAMMER-TECHNION e CITATION e as temperaturas máximas atingidas nos alvos foram calculadas com o código MTRCR-IEAR1. Os cálculos demonstram que a irradiação do dispositivo deverá ocorrer sem consequências adversas à operação do reator. A quantidade total de 99Mo foi calculada com o programa SCALE e considerando que o tempo necessário para o processamento químico e recuperação do 99Mo será de cinco dias após a irradiação, teremos disponível para distribuição uma atividade de 99Mo de 176 Ci para 3 dias de irradiação, 236 Ci para 5 dias de irradiação e 272 Ci para 7 dias de irradiação dos alvos. / Technetium-99m (99mTc), the product of radioactive decay of molybdenum-99 (99Mo), is one of the most widely used radioisotope in nuclear medicine, covering approximately 80% of all radiodiagnosis procedures in the world. Nowadays, Brazil requires an amount of about 450 Ci of 99Mo per week. Due to the crisis and the shortage of 99Mo supply chain that has been observed on the world since 2008, IPEN/CNEN-SP decided to develop a project to produce 99Mo through fission of uranium-235. The objective of this dissertation was the development of neutronic and thermal-hydraulic calculations to evaluate the operational safety of a device for 99Mo production to be irradiated in the IEA-R1 reactor core at 5 MW. In this device will be placed ten targets of UAlx-Al dispersion fuel with low enriched uranium (LEU) and density of 2.889 gU/cm³. For the neutronic calculations were utilized the computer codes HAMMER-TECHNION and CITATION and the maximum temperatures reached in the targets were calculated with the code MTRCR-IEAR1. The analysis demonstrated that the device irradiation will occur without adverse consequences to the operation of the reactor. The total amount of 99Mo was calculated with the program SCALE and considering that the time needed for the chemical processing and recovering of the 99Mo will be five days after the irradiation, we have that the 99Mo activity available for distribution will be 176 Ci for 3 days of irradiation, 236 Ci for 5 days of irradiation and 272 Ci for 7 days of targets irradiation.

O presente estudo foi realizado para verificar a possibilidade de redução do núcleo do reator IEA-R1 do IPEN/CNEN-SP. Cálculos neutrônicos foram desenvolvidos para um conjunto de novas configurações para que, a posteriori, a análise termo-hidráulica e de segurança pudessem ser realizadas. As novas configurações analisadas são menores por diversos motivos, como obter uma melhor utilização do combustível, melhor distribuição dos fluxos de nêutrons, dentre outros. Para que se possa atingir tais configurações, a densidade de Urânio no combustível deve ser aumentada. Neste estudo, combustíveis de U3Si2-Al com 4,8gU/cm3 foram testados e novos núcleos para o reator IEA-R1 foram propostos e discutidos. A análise neutrônica não impõe restrições aos núcleos estudados. A análise termohidráulica mostrou que as margens de segurança e os perfis de temperatura ao longo das placas combustíveis não excedem os limites de projeto. Os coeficientes de temperatura obtidos para os novos núcleos, no caso isotérmico, são todos negativos, conforme desejado. A queima mostrou que núcleos supercompactos não apresentam excesso de reatividade suficiente para o funcionamento dos mesmo, ao se utilizar combustíveis com 4,8gU/cm3. Um APR (Acidente de Perda de Refrigerante) foi simulado para os núcleos remanescentes. A ruptura da fronteira do primário se mostrou o acidente mais crítico, devido ao curto tempo para o esvaziamento completo da piscina do reator. As temperaturas atingidas após o descobrimento foram calculadas e não excedem aquelas cujos valores propiciam empolamento nas placas combustíveis (475 °! a 550 °!), uma vez que se obedeça os tempos de esvaziamento seguro da piscina para as novas configurações. / This study was performed considering prospective candidates for the IPEN/CNEN-SP IEA-R1 research reactor core. Some neutronic calculations were developed for a set of new core configurations to push forward the thermal-hydraulic and safety analysis. The new core configurations will be smaller for several reasons (e.g., better fuel utilization, neutron fluxes and so on). To achieve such smaller arrangements, the U-fuel density has to be increased. In the current study, configurations with 4.8gU/cm3 U3Si2- Al fuels were tested using the software MCNP and a set of new core configurations for the IPEN/CNEN-SP IEA-R1 research reactor has been presented and discussed. The Neutronic analysis imposes no restrictions on the new cores. The Thermal- Hydraulic (TH) analysis showed that the safety margins and the temperature profile through the fuel plate dont exceed the design limits. The isothermal temperature coefficients were calculated being all negative, as desired. The burnup concludes that super compact cores dont have enough excess reactivity to keep the reactor working with 4.8gU/cm3 U3Si2-Al fuels. A LOCA (Loss of Cooling Accident) was simulated for the remaining cores. The border rupture of the primary system was the most critical accident, due to the short time for the complete emptying of the reactor pool. The temperatures reached after this accident were calculated and dont exceed the fuel plates limits (475 °C - 550 °C), once the time for safe emptying are taken into account for the IEA-R1 pool.

Ce travail traite de la thermohydraulique d’un bain de melt couplée à la physicochimie pour ladescription du comportement de mélanges de matériaux (non-eutectiques).On décrit le transitoire d’établissement de température dans un liquide avec dégagement de puissancevolumique en présence de solidification sur une paroi refroidie. Le modèle développé à cet effet estvalidé par rapport aux résultats des essais LIVE réalisés à KIT. Dans les conditions de ces essais onmontre que la température d’interface suit la température liquidus (correspondant à la composition dubain liquide) pendant le transitoire d’établissement de la température dans le bain et des croûtessolides.Par ailleurs, on propose un modèle d’interaction entre un liquide non-eutectique (soumis à dissipationvolumique de puissance) et une paroi fusible dont la température de fusion est inférieure à latempérature liquidus du bain. Les prédictions du modèle sont comparées aux résultats des essaisARTEMIS 2D. On en déduit une nouvelle formulation de la température d’interface (inférieure àliquidus température) entre le liquide et la couche pâteuse en paroi. / This work deals with the thermal-hydraulics of a melt pool coupled with the physical chemistry for thepurpose of describing the behaviour of mixtures of materials (non-eutectic).Evolution of transient temperature in a liquid melt pool heated by volumetric power dissipation hasbeen described with solidification on the cooled wall. The model has been developed and is validatedfor the experimental results given by LIVE experiment, performed at Karlsruhe Institute ofTechnology (KIT) in Germany. Under the conditions of these tests, it is shown that the interfacetemperature follows the liquidus temperature (corresponding to the composition of the liquid bath)during the whole transient. Assumption of interface temperature as liquidus temperature allowsrecalculating the evolution of the maximum melt temperature as well as the local crust thickness.Furthermore, we propose a model for describing the interaction between a non-eutectic liquid meltpool (subjected to volumetric power dissipation) and an ablated wall whose melting point is below theliquidus temperature of the melt. The model predictions are compared with results of ARTEMIS 2Dtests. A new formulation of the interface temperature between the liquid melt and the solid wall(below liquidus temperature) has been proposed.

O presente estudo foi realizado para verificar a possibilidade de redução do núcleo do reator IEA-R1 do IPEN/CNEN-SP. Cálculos neutrônicos foram desenvolvidos para um conjunto de novas configurações para que, a posteriori, a análise termo-hidráulica e de segurança pudessem ser realizadas. As novas configurações analisadas são menores por diversos motivos, como obter uma melhor utilização do combustível, melhor distribuição dos fluxos de nêutrons, dentre outros. Para que se possa atingir tais configurações, a densidade de Urânio no combustível deve ser aumentada. Neste estudo, combustíveis de U3Si2-Al com 4,8gU/cm3 foram testados e novos núcleos para o reator IEA-R1 foram propostos e discutidos. A análise neutrônica não impõe restrições aos núcleos estudados. A análise termohidráulica mostrou que as margens de segurança e os perfis de temperatura ao longo das placas combustíveis não excedem os limites de projeto. Os coeficientes de temperatura obtidos para os novos núcleos, no caso isotérmico, são todos negativos, conforme desejado. A queima mostrou que núcleos supercompactos não apresentam excesso de reatividade suficiente para o funcionamento dos mesmo, ao se utilizar combustíveis com 4,8gU/cm3. Um APR (Acidente de Perda de Refrigerante) foi simulado para os núcleos remanescentes. A ruptura da fronteira do primário se mostrou o acidente mais crítico, devido ao curto tempo para o esvaziamento completo da piscina do reator. As temperaturas atingidas após o descobrimento foram calculadas e não excedem aquelas cujos valores propiciam empolamento nas placas combustíveis (475 °! a 550 °!), uma vez que se obedeça os tempos de esvaziamento seguro da piscina para as novas configurações. / This study was performed considering prospective candidates for the IPEN/CNEN-SP IEA-R1 research reactor core. Some neutronic calculations were developed for a set of new core configurations to push forward the thermal-hydraulic and safety analysis. The new core configurations will be smaller for several reasons (e.g., better fuel utilization, neutron fluxes and so on). To achieve such smaller arrangements, the U-fuel density has to be increased. In the current study, configurations with 4.8gU/cm3 U3Si2- Al fuels were tested using the software MCNP and a set of new core configurations for the IPEN/CNEN-SP IEA-R1 research reactor has been presented and discussed. The Neutronic analysis imposes no restrictions on the new cores. The Thermal- Hydraulic (TH) analysis showed that the safety margins and the temperature profile through the fuel plate dont exceed the design limits. The isothermal temperature coefficients were calculated being all negative, as desired. The burnup concludes that super compact cores dont have enough excess reactivity to keep the reactor working with 4.8gU/cm3 U3Si2-Al fuels. A LOCA (Loss of Cooling Accident) was simulated for the remaining cores. The border rupture of the primary system was the most critical accident, due to the short time for the complete emptying of the reactor pool. The temperatures reached after this accident were calculated and dont exceed the fuel plates limits (475 °C - 550 °C), once the time for safe emptying are taken into account for the IEA-R1 pool.

Tecnécio-99m (99mTc), o produto de decaimento do molibdênio-99 (99Mo), é um dos radioisótopos mais utilizados na medicina nuclear, abrangendo cerca de 80% de todos os procedimentos de radiodiagnóstico médico pelo mundo. Atualmente o Brasil necessita de uma quantidade de aproximadamente 450 Ci de 99Mo por semana. Devido à crise e à escassez em seu fornecimento que vem sendo observada no cenário mundial desde 2008, o IPEN decidiu desenvolver um projeto próprio para produção de 99Mo através da fissão do urânio-235. O objetivo deste trabalho de dissertação foi desenvolver cálculos neutrônicos e temo-hidráulicos para avaliar a segurança operacional de um dispositivo para produção de 99Mo a ser irradiado no núcleo do reator IEA-R1. Neste dispositivo serão alojados dez alvos do tipo dispersão de UAlx-Al com baixo enriquecimento de urânio (LEU) e densidade de 2,889 gU/cm³. Para o cálculo neutrônico foram utilizados os programas computacionais HAMMER-TECHNION e CITATION e as temperaturas máximas atingidas nos alvos foram calculadas com o código MTRCR-IEAR1. Os cálculos demonstram que a irradiação do dispositivo deverá ocorrer sem consequências adversas à operação do reator. A quantidade total de 99Mo foi calculada com o programa SCALE e considerando que o tempo necessário para o processamento químico e recuperação do 99Mo será de cinco dias após a irradiação, teremos disponível para distribuição uma atividade de 99Mo de 176 Ci para 3 dias de irradiação, 236 Ci para 5 dias de irradiação e 272 Ci para 7 dias de irradiação dos alvos. / Technetium-99m (99mTc), the product of radioactive decay of molybdenum-99 (99Mo), is one of the most widely used radioisotope in nuclear medicine, covering approximately 80% of all radiodiagnosis procedures in the world. Nowadays, Brazil requires an amount of about 450 Ci of 99Mo per week. Due to the crisis and the shortage of 99Mo supply chain that has been observed on the world since 2008, IPEN/CNEN-SP decided to develop a project to produce 99Mo through fission of uranium-235. The objective of this dissertation was the development of neutronic and thermal-hydraulic calculations to evaluate the operational safety of a device for 99Mo production to be irradiated in the IEA-R1 reactor core at 5 MW. In this device will be placed ten targets of UAlx-Al dispersion fuel with low enriched uranium (LEU) and density of 2.889 gU/cm³. For the neutronic calculations were utilized the computer codes HAMMER-TECHNION and CITATION and the maximum temperatures reached in the targets were calculated with the code MTRCR-IEAR1. The analysis demonstrated that the device irradiation will occur without adverse consequences to the operation of the reactor. The total amount of 99Mo was calculated with the program SCALE and considering that the time needed for the chemical processing and recovering of the 99Mo will be five days after the irradiation, we have that the 99Mo activity available for distribution will be 176 Ci for 3 days of irradiation, 236 Ci for 5 days of irradiation and 272 Ci for 7 days of targets irradiation.

Les composants du circuit primaire des réacteurs nucléaires à eau sous pression (REP) subissent une corrosion généralisée entraînant le relâchement d'espèces solubles dans le fluide primaire (principalement Fe, Ni, Cr, Mn, Co). Sous l'effet de la convection du fluide, ces espèces sont entraînées dans le circuit primaire. Une partie de ces espèces peut précipiter sur les surfaces du combustible et être activée sous l'effet du flux neutronique régnant dans cette région. Ce dépôt de produits de corrosion peut, sous l'effet des forces hydrodynamiques du fluide primaire, être érodé (ou bien dissous si les conditions thermo-chimiques le permettent). Ces espèces activées (principalement du 58Co, 60Co, 51Cr et 54Mn), sous l'effet de la convection vont se retrouver disséminées dans l'ensemble du circuit primaire où elles pourront se redéposer (ou bien précipiter) sur les différents composants et ainsi contaminer l'ensemble du circuit primaire. Au cours d'un cycle de fonctionnement normal dans un REP EDF, l'activité du fluide dans le circuit primaire est relativement constante (généralement de l'ordre de 10-20 MBq.m-3 en 58Co). Cependant, lors de certains cycles de fonctionnement (en fonction de la gestion de combustible), notamment on observe des montées d'activités volumiques importantes en 58Co et en 51Cr pouvant atteindre une centaine de fois celles observées habituellement. Ces montées d'activités volumiques sont dues à l'établissement dans les régions les plus "chaudes" des assemblages de combustible d'un régime d'ébullition nucléée. L'ébullition peut dans certains cas multiplier par un facteur 10 à 100 l'épaisseur de dépôt formé sur le combustible conduisant ainsi à un transfert de masse plus important sous forme particulaire entre le dépôt et le fluide primaire du fait de l'érosion. Une modélisation des mécanismes de transfert de masse entre le fluide primaire et le dépôt sur ces régions "chaudes" du combustible en régime d'ébullition nucléée et les impacts sur la contamination du circuit primaire sont décrits dans ce mémoire. L'ébullition à la surface du dépôt ou bien dans le dépôt lui-même provoque un enrichissement à la paroi en espèces ioniques pouvant entraîner une précipitation plus importante ou bien modifier le comportement d'une espèce d'un régime de dissolution à un régime de précipitation ; le dépôt de particules turbulent et inertiel est lui aussi favorisé. La vaporisation du fluide à la paroi ainsi que la formation des bulles elles-mêmes entraînent aussi un dépôt et une précipitation plus importants. La prise en compte de ces mécanismes de transfert de masse dans le code OSCAR (Outil de Simulation de la ContAmination en Réacteur), développé au sein du Laboratoire de Modélisation des interactions et Transferts en Réacteur au CEA, conduit à une bonne reproduction des résultats expérimentaux issus du retour d'expérience des centrales françaises tant au niveau des dépôts formés dans les régions avec ébullition que des activités volumiques. / The Pressurized Water Reactors (PWRs) primary circuit materials are subject to general corrosion leading to soluble metallic element (mainly Fe, Ni, Cr, Mn, Co) transfer and subsequent ion precipitation processes on the primary circuit surfaces. When deposited on fuel rods, these species are activated by neutron flux. Thus, crud erosion and dissolution processes induce to primary coolant activity. During a normal operating cycle in a EDF PWR, the volume activity in the coolant is relativly stable (usually about 10-20 MBq.m−3 in 58Co). In some cycles (depending on fuel management), significant increases in 58Co and 51Cr volume activities are observed (10 to 100 times the ordinary volume activities). These increases of volume activities are due to local sub-cooled nucleate boiling on the "hot" parts of fuel assemblies. As presented in this thesis, boiling at the top of some fuel assemblies may lead to much higher amount of metallic elements than usual (some micrometers). Indeed, boiling that can locally occurs under PWR conditions concentrates species and to increase significantly the quantity of deposited and precipitated material. Erosion flux is higher in these regions due to thicker crud thickness, involving a greater mass transfer of activated isotopes to the primary coolant. The OSCAR calculation code, developed by the "Laboratoire de Modélisation des interactions et Transferts en Réacteur" in CEA, with these new mass transfer models can now well estimate the amount of deposit and the volume activities in the primary coolant in case of boiling in accordance with french PWR measurements.

Un nouveau procédé de trigénération thermo-hydraulique fonctionnant à partir d'énergie thermique basse température (80 à 110 °C) a été étudié pour assurer les différents besoins du secteur résidentiel. Le terme "thermo-hydraulique" se réfère à l'utilisation d'un liquide incompressible qui permet de transférer le travail hydrauliquement entre différents composants ou sous-systèmes, permettant d'améliorer l'efficacité de la chaine de conversion énergétique. Un modèle quasi-statique a été développé pour évaluer les performances énergétiques des différentes variantes du procédé. Ces calculs ont permis de définir parmi un large choix, quels fluides de travail étaient les plus appropriés. Ces calculs ont été complétés par une étude quasi-dynamique et dynamique permettant un meilleur dimensionnement du procédé. Enfin, une étude de fonctionnement annuel a été réalisée à partir du modèle quasi-statique pour évaluer l'évolution des performances ainsi que sa production d'énergie sur une année complète de fonctionnement. Ces études montrent que le couple fluide R1234yf/R1233zd semble le plus approprié à un fonctionnement en climat méditerranéen. L'étude annuelle montre qu'en considérant les données climatiques de la ville de Perpignan, le procédé permet d'amplifier l'énergie solaire collectée d'un facteur de 1,32 en moyenne et permet d'atteindre un COP solaire de 0,24 en mode rafraichissement. Quand les besoins thermiques sont satisfaits, l'intégralité de l'énergie solaire captée est valorisée pour produire de l'électricité avec un rendement moyen annuel de 4,2%. / A new process based on thermal-hydraulic conversion actuated by low-grade thermal energy (80–110 °C) is investigated and aims at providing trigeneration energy features for the residential sector. "Thermo-hydraulic" term refers to a process involving an incompressible fluid used as an intermediate medium to transfer work hydraulically between different thermal operated components or sub-systems allowing to improve the efficiency of the energy conversion chain. A model, assuming steady-state operations, is developed to assess the energy performances of different variants of this thermo-hydraulic process as well as various pairs of working fluids. These calculations were completed by a quasi-dynamic and dynamic models allowing a better sizing of the process. Finally, an annual study was realized from the quasi-static model in order to estimate the evolution of the performances as well as its power production over a complete year of functioning. For instance, in the frame of a single-family home, located in the Mediterranean region, the working fluid pair (R1234yf/R1233zd) is investigated in detail in order to estimate the annual performances. For domestic houses, the process aims at amplifying the solar energy collected by a factor of 1.32 for heating purpose, provides a cold production with a solar COP of 0.24 and generates electricity from the remaining solar energy with an efficiency of 4.2%.