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1
MAI, LUIZ A. "Sistema de obtencao de um pre-projeto otimizado de um nucleo de um reator nuclear." reponame:Repositório Institucional do IPEN, 1988. http://repositorio.ipen.br:8080/xmlui/handle/123456789/9914.
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Dissertacao (Mestrado)
IPEN/D
Instituto de Pesquisas Energeticas e Nucleares - IPEN/CNEN-SP
2
HIROMOTO, MARIA Y. K. "PSINCO-um programa para calculo da distribuicao de potencia e supervisao do nucleo de reatores nucleares, utilizando sinais de detetores tipo 'SPD'." reponame:Repositório Institucional do IPEN, 1998. http://repositorio.ipen.br:8080/xmlui/handle/123456789/10706.
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Dissertacao (Mestrado)
IPEN/D
Instituto de Pesquisas Energeticas e Nucleares - IPEN/CNEN-SP
3
CARVALHO, LUIZ S. "Frequencia de danos no nucleo por blecaute em reator nuclear de concepcao avancada." reponame:Repositório Institucional do IPEN, 2004. http://repositorio.ipen.br:8080/xmlui/handle/123456789/11147.
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Dissertacao (Mestrado)
IPEN/D
Instituto de Pesquisas Energeticas e Nucleares - IPEN/CNEN-SP
4
Laufer, Michael Robert. "Granular Dynamics in Pebble Bed Reactor Cores." Thesis, University of California, Berkeley, 2013. http://pqdtopen.proquest.com/#viewpdf?dispub=3593891.
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This study focused on developing a better understanding of granular dynamics in pebble bed reactor cores through experimental work and computer simulations. The work completed includes analysis of pebble motion data from three scaled experiments based on the annular core of the Pebble Bed Fluoride Salt-Cooled High- Temperature Reactor (PB-FHR). The experiments are accompanied by the development of a new discrete element simulation code, GRECO, which is designed to offer a simple user interface and simplified two-dimensional system that can be used for iterative purposes in the preliminary phases of core design. The results of this study are focused on the PB-FHR, but can easily be extended for gas-cooled reactor designs.
Experimental results are presented for three Pebble Recirculation Experiments (PREX). PREX 2 and 3.0 are conventional gravity-dominated granular systems based on the annular PB-FHR core design for a 900 MWth commercial prototype plant and a 16 MWth test reactor, respectively. Detailed results are presented for the pebble velocity field, mixing at the radial zone interfaces, and pebble residence times. A new Monte Carlo algorithm was developed to study the residence time distributions of pebbles in different radial zones. These dry experiments demonstrated the basic viability of radial pebble zoning in cores with diverging geometry before pebbles reach the active core.
Results are also presented from PREX 3.1, a scaled facility that uses simulant materials to evaluate the impact of coupled fluid drag forces on the granular dynamics in the PB-FHR core. PREX 3.1 was used to collect first of a kind pebble motion data in a multidimensional porous media flow field. Pebble motion data were collected for a range of axial and cross fluid flow configurations where the drag forces range from half the buoyancy force up to ten times greater than the buoyancy force. Detailed analysis is presented for the pebble velocity field, mixing behavior, and residence time distributions for each fluid flow configuration.
The axial flow configurations in PREX 3.1 showed small changes in pebble motion compared to a reference case with no fluid flow and showed similar overall behavior to PREX 3.0. This suggests that dry experiments can be used for core designs with uniform one-dimensional coolant flow early in the design process at greatly reduced cost. Significant differences in pebble residence times were observed in the cross fluid flow configurations, but these were not accompanied by an overall horizontal diffusion bias. Radial zones showed only a small shift in position due to mixing in the diverging region and remained stable in the active core. The results from this study support the overall viability of the annular PB-FHR core by demonstrating consistent granular flow behavior in the presence of complex reflector geometries and multidimensional fluid flow fields.
GRECO simulations were performed for each of the experiments in this study in order to develop a preliminary validation basis and to understand for which applications the code can provide useful analysis. Overall, the GRECO simulation results showed excellent agreement with the gravity-dominated PREX experiments. Local velocity errors were found to be generally within 10-15% of the experimental data. Average radial zone interface positions were predicted within two pebble diameters. GRECO simulations over predicted the amount of mixing around the average radial zone interface position and therefore can be treated as a conservative upper bound when used in neutronics analysis. Residence time distributions from the GRECO velocity data based on the Monte Carlo algorithm closely matched those derived from the experiment velocity statistics. GRECO simulation results for PREX 3.1 with coupled drag forces showed larger errors compared to the experimental data, particularly in the cases with cross fluid flow. The large discrepancies suggest that GRECO results in systems with coupled fluid drag forces cannot be used with high confidence at this point and future development work on coupled pebble and fluid dynamics with multidimensional fluid flow fields is required.
5
Jahn, Gordon James. "Agent-based structural condition monitoring for nuclear reactor cores." Thesis, University of Strathclyde, 2011. http://oleg.lib.strath.ac.uk:80/R/?func=dbin-jump-full&object_id=17400.
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A significant proportion of the UK energy needs are currently serviced by a fleet of ageing nuclear reactors. Ensuring that these reactors are operated safely is the highest priority and the structural health of their cores, that provide channels for control rods and coolant gas, is a key aspect. This thesis focuses on the application of structuralhealth monitoring to the graphite reactor cores used in the UK and presents a specification for the use of structural health monitoring (SHM) techniques already es- tablished in bridge and aircraft monitoring, with data obtained through existing reactor monitoring processes. This approach utilises statistical and clustering techniques on monitoring data that can be acquired during online operation of the plant. The use of existing monitoring processes to complement the established inspection regime for nuclear reactors is a novel contribution from this work. As part of proving the SHM approach, this thesis reports on work undertaken to identify suitable data and numerical limits for the cluster analysis. This analysis considers the data with respect to the stated aim of detectin~ core distortion and demonstrates that the chosen data and values are acceptable and conservative in the context of reactor condition monitoring. An assessment of the SHM solution is presented describing the im- plementation of the SHM approach using a multi-agent system (MAS), IMAPS. This implementation required consideration of using MAS tech- nology for condition monitoring, and the novel contribution of a technique for storing and retrieving historical data in a manner concomitant with both MAS and relational database theory is presented.ij The thesis concludes that condition monitoring is feasible on the graphite cores, and that multi-variate analysis through SHM implemented within a MAS offers a storage and analysis platform that can both handle the data volumes and accommodate further extensions as required.
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Shuffler, Carter Alexander. "Optimization of hydride fueled pressurized water reactor cores." Thesis, Massachusetts Institute of Technology, 2004. http://hdl.handle.net/1721.1/33634.
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Thesis (S.M.)--Massachusetts Institute of Technology, Dept. of Nuclear Engineering, 2004.Includes bibliographical references (leaf 173).
This thesis contributes to the Hydride Fuels Project, a collaborative effort between UC Berkeley and MIT aimed at investigating the potential benefits of hydride fuel use in light water reactors (LWRs). This pursuit involves implementing an appropriate methodology for design and optimization of hydride and oxide fueled cores. Core design is accomplished for a range of geometries via steady-state and transient thermal hydraulic analyses, which yield the maximum power, and fuel performance and neutronics studies, which provide the achievable discharge burnup. The final optimization integrates the outputs from these separate studies into an economics model to identify geometries offering the lowest cost of electricity, and provide a fair basis for comparing the performance of hydride and oxide fuels. Considerable work has already been accomplished on the project; this thesis builds on this previous work. More specifically, it focuses on the steady-state thermal hydraulic and economic analyses for pressurized water reactor (PWR) cores utilizing UZrH₁.₆ and UO₂. A previous MIT study established the steady-state thermal hydraulic design methodology for determining maximum power from square array PWR core designs.
(cont.) The analysis was not performed for hexagonal arrays under the assumption that the maximum achievable powers for both configurations are the same for matching rod diameters and H/HM ratios. This assumption is examined and verified in this work by comparing the thermal hydraulic performance of a single hexagonal core with its equivalent square counterpart. In lieu of a detailed vibrations analysis, the steady-state thermal hydraulic analysis imposed a single design limit on the axial flow velocity. The wide range of core geometries considered and the large power increases reported by the study makes it prudent to refine this single limit approach. This work accomplishes this by developing and incorporating additional design limits into the thermal hydraulic analysis to prevent excessive rod vibration and wear. The vibrations and wear mechanisms considered are: vortex-induced vibration, fluid-elastic instability, turbulence-induced vibration, fretting wear, and sliding wear. Concomitantly with this work, students at UC Berkeley and MIT have undertaken the neutronics, fuel performance, and transient thermal hydraulic studies.
(cont.) With these results, and the output from the steady-state thermal hydraulic analysis with vibrations and wear imposed design limits, an economics model is employed to determine the optimal geometries for incorporation into existing PWRs. The model also provides a basis for comparing the performance of UZrH₁.₆ to UO₂ for a range of core geometries. Though this analysis focuses only on these fuels, the methodology can easily be extended to additional hydride and oxide fuel types, and will be in the future. Results presented herein do not show significant cost savings for UZrH₁.₆, primarily because the power and energy generation per core loading for both fuels are similar. Furthermore, the most economic geometries typically do not occur where power increases are reported by the thermal hydraulics. As a final note, the economic results in this report require revision to account for recent changes in the fuel performance analysis methodology. The changes, however, are not expected to influence the overall conclusion that UZrH₁.₆ does not outperform UO₂ economically.
by Carter Alexander Shuffler.
S.M.
7
Trant, Jarrod Michael. "Transient analysis of hydride fueled pressurized water reactor cores." Thesis, Massachusetts Institute of Technology, 2004. http://hdl.handle.net/1721.1/33632.
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Thesis (S.M.)--Massachusetts Institute of Technology, Dept. of Nuclear Engineering, 2004.Includes bibliographical references (leaves 132-133).
This thesis contributes to the hydride nuclear fuel project led by U. C. Berkeley for which MIT is to perform the thermal hydraulic and economic analyses. A parametric study has been performed to determine the optimum combination of lattice pitch, rod diameter, and channel shape-further referred to as geometry-for maximizing power given specific transient conditions for pressurized water reactors (PWR) loaded with either U02 or UZrH1.6 fuel. Several geometries have been examined with the VIPRE subchannel analysis tool along with MATLAB scripts previously developed to automate VIPRE execution. The transients investigated were a large break loss of coolant accident (LBLOCA), am overpower transient, and a complete loss of flow accident. The maximum achievable power for each geometry is defined as the highest power that can be sustained without exceeding any of the steady state or transient limits. The limits were chosen based on technical feasibility and safety of the reference core and compared with the final safely analysis report (FSAR) of the reference core, the South Texas Project Electric Generating Station (STPEGS), whenever possible. This analysis was performed for two separate pressure drop limits of 29 and 60 psia for both a square array with grid spacers and a hexagonal array with wire wraps.
(cont.) The square core geometry sustaining the highest power (4820.0 MW) for both the hydride and oxide fueled has a pitch of 9.0 mm and a rod diameter of 6.5 mm and was limited by the complete loss of flow accident. Both of these maximum power geometries occurred at the 60 psia pressure drop case. The maximum power of the 29 psia pressure drop case (4103.9 MW) for both fuel types occurred at a pitch of 9.7 mm and a rod diameter of 6.5 mm. The maximum power for the hexagonal arrayed cores occurred at the same hydrogen to heavy metal ratio as the square cores. The hydride fueled core power (5123.2 MW) was limited by the overpower transient while the oxide fueled core power (4996.1 MW) was limited by the overpower transient. The pressure drop constraint was not limiting for either fuel type for either pressure drop case for the wire wrapped cores.
by Jarrod Michael Trant.
S.M.
8
Alam, Syed Bahauddin. "The design of reactor cores for civil nuclear marine propulsion." Thesis, University of Cambridge, 2018. https://www.repository.cam.ac.uk/handle/1810/275650.
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Perhaps surprisingly, the largest experience in operating nuclear power plants has been in nuclear naval propulsion, particularly submarines. This accumulated experience may become the basis of a proposed new generation of compact nuclear power plant designs. In an effort to de-carbonise commercial freight shipping, there is growing interest in the possibility of using nuclear propulsion systems. Reactor cores for such an application would need to be fundamentally different from land-based power generation systems, which require regular refueling, and from reactors used in military submarines, as the fuel used could not conceivably be as highly enriched. Nuclear-powered propulsion would allow ships to operate with low fuel costs, long refueling intervals, and minimal emissions; however, currently such systems remain largely confined to military vessels. This research project undertakes computational modeling of possible soluble-boron-free (SBF) reactor core designs for this application, with a view to informing design decisions in terms of choices of fuel composition, materials, core geometry and layout. Computational modeling using appropriate reactor physics (e.g. WIMS, MONK, Serpent and PANTHER), thermal-hydraulics etc. codes (e.g. COBRA-EN) is used for this project. With an emphasis on reactor physics, this study investigates possible fuel assembly and core designs for civil marine propulsion applications. In particular, it explores the feasibility of using uranium/thorium-rich fuel in a compact, long-life reactor and seek optimal choices and designs of the fuel composition, reactivity control, assembly geometry, and core loading in order to meet the operational needs of a marine propulsion reactor. In this reactor physics and 3D coupled neutronics/thermal-hydraulics study, we attempt to design a civil marine reactor core that fulfills the objective of providing at least 15 effective full-power-years (EFPY) life at 333 MWth. In order to unleash the benefit of thorium in a long life core, the micro-heterogeneous ThO2-UO2 duplex fuel is well-positioned to be utilized in our proposed civil marine core. Unfortunately, A limited number of studies of duplex fuel are available in the public domain, but its use has never been examined in the context of a SBF environment for long-life small modular rector (SMR) core. Therefore, we assumed micro-heterogeneous ThO2-UO2 duplex fuel for our proposed marine core in order to explore its capability. For the proposed civil marine propulsion core design, this study uses 18% U-235 enriched micro-heterogeneous ThO2-UO2 duplex fuel. To provide a basis for comparison we also evaluate the performance of homogeneously mixed 15% U-235 enriched all-UO2 fuel. This research also attempts to design a high power density core with 14 EFPY while satisfying the neutronic and thermal-hydraulics safety constraints. A core with an average power density of 100 MW/m3 has been successfully designed while obtaining a core life of 14 years. The average core power density for this core is increased by ∼50% compared to the reference core design (63 MW/m3 and is equivalent to Sizewell B PWR (101.6 MW/m3 which means capital costs could be significantly reduced and the economic attractiveness of the marine core commensurately improved. In addition, similar to the standard SMR core, a reference core with a power density of 63 MW/m3 has been successfully designed while obtaining a core life of ∼16 years. One of the most important points that can be drawn from these studies is that a duplex fuel lattice needs less burnable absorber than uranium-only fuel to achieve the same poison performance. The higher initial reactivity suppression and relatively smaller reactivity swing of the duplex can make the task of reactivity control through BP design in a thorium-rich core easier. It is also apparent that control rods have greater worth in a duplex core, reducing the control material requirements and thus potentially the cost of the rods. This research also analyzed the feasibility of using thorium-based duplex fuel in different cases and environments to observe whether this fuel consistently exhibit superior performance compared to the UO2 core in both the assembly and whole-core levels. The duplex fuel/core consistently exhibits superior performance in consideration of all the neutronic and TH constraints specified. It can therefore be concluded from this study that the superior performance of the thorium-based micro-heterogeneous ThO2-UO2 duplex fuel provides enhanced confidence that this fuel can be reliably used in high power density and long-life SBF marine propulsion core systems, offering neutronic advantages compared to the all-UO2 fuel. Last, but not least, considering all these factors, duplex fuel can potentially open the avenue for low-enriched uranium (LEU) SBF cores with different configurations. Motivated by growing environmental concerns and anticipated economic pressures, the overall goal of this study is to examine the technological feasibility of expanding the use of nuclear propulsion to civilian maritime shipping and to identify and propose promising candidate core designs.
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PINTO, LETICIA N. "Experimentos de efeitos de reatividade no reator nuclear IPEN/MB-01." reponame:Repositório Institucional do IPEN, 2012. http://repositorio.ipen.br:8080/xmlui/handle/123456789/10099.
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Dissertação (Mestrado)
IPEN/D
Instituto de Pesquisas Energeticas e Nucleares - IPEN-CNEN/SP
10
SANTOS, DIOGO F. dos. "Caracterização dos campos neutrônicos obtidos por meio de armadilhas de nêutrons a partir da utilização de água pesada (D2O) no interior do núcleo do reator nuclear IPEN/MB-01." reponame:Repositório Institucional do IPEN, 2015. http://repositorio.ipen.br:8080/xmlui/handle/123456789/23825.
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Dissertação (Mestrado em Tecnologia Nuclear)
IPEN/D
Instituto de Pesquisas Energeticas e Nucleares - IPEN-CNEN/SP
11
Wallace, Christopher John. "Distributed data fusion for condition monitoring of graphite nuclear reactor cores." Thesis, University of Strathclyde, 2013. http://oleg.lib.strath.ac.uk:80/R/?func=dbin-jump-full&object_id=20607.
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Nuclear power stations worldwide are exceeding their originally specified design lives and with only limited construction of new generation underway, there is a desire to continue the operation of existing stations to ensure electricity supply. Continued operation of nuclear power stations with degrading and life-limiting components necessitates increased monitoring and inspection, particularly of the reactor cores, to ensure they are safe to operate. The monitoring of a large number of components and their related data sources is a distributed and time consuming process for the engineer given the lack of infrastructure available for collecting, managing and analysing monitoring data. This thesis describes the issues associated with nuclear Condition Monitoring (CM) and investigates the suitability of a distributed framework utilising intelligent software agents to collect, manage and analyse data autonomously. The application of data fusion techniques is examined to estimate unre corded parameters, provide contextualisation for anomalies in order to quickly identify true faults from explainable anomalies and to extract more detail from existing CM data. A generalised framework is described for nuclear CM of any type of reactor, specifying the required components and capabilites based on the design of a suitable Multi Agent System, including the interaction of the framework with existing CM systems and human users. A high level ontology for nuclear CM is proposed and is emphasised as a crucial aspect of the data management and extendability of the framework to incorporate further data sources and analyses. A prototype system, based on the generalised framework is developed for the case of the Advanced Gas-cooled Reactor, with new and existing CM analyses formalised within intelligent agents. Using real station data and simulated fault data, the prototype system was shown to be capable of performing the existing monitoring tasks considerably faster than a human user while retaining all data and analyses for justification and traceability of decisions based on the analyses.
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Inzerillo, Santo. "Nonlinear estimation for condition monitoring of advanced gas-cooled nuclear reactor cores." Thesis, University of Strathclyde, 2012. http://oleg.lib.strath.ac.uk:80/R/?func=dbin-jump-full&object_id=19546.
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As the Advanced Gas-cooled Reactor (AGR) nuclear power stations are ageing, the nuclear core composed by graphite bricks can distort. The direct measurement of the core condition is costly and time-consuming, hence, alternative methods have been developed to provide the necessary information about the core condition. This thesis presents a model-based technique for condition monitoring of AGRs cores using measurements obtained during routine core refuelling process. It has been demonstrated that Fuel Grab Load Trace (FGLT) data gathered during refuelling operations provides, through the magnitude of its friction component, information relating to the condition of the graphite bricks. Therefore, the condition monitoring of an AGR leads to the estimation of the friction force resulting from the interaction of the fuel assembly and the core channel. To this end the main objective of this work is to investigate estimation techniques that are needed in industrial applications and in particular can be used in the refuelling filtering problem. As a result of this study, a novel LPV estimator and robust estimator have been designed and implemented. A model for the refuelling system was initially developed from the first principles of the process. Then its fuel assembly dynamics subsystem was identified to be used in a model based filtering application. Finally a H robust estimator was employed to estimate the friction force to be used for the core condition analysis.
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Redd, Evan M. "X-10 reactor forensic analysis and evaluation using a suite of neutron transport codes." Thesis, Georgia Institute of Technology, 2015. http://hdl.handle.net/1853/53978.
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X-10, the genesis production reactor for the U.S. paved the way for all weapons material production. This feat offers a unique fundamental opportunity of nuclear forensic analysis and popular neutron code package evaluation. Production reactor nuclear forensic signatures and characteristics are emphasized throughout this work. These underlying production characteristics are reported and analyzed for potential in-core zone provenance and axial slug location coupled with how the nuclear data uncertainties affect these conclusions. Material attribution with respect to commercial versus military reactor applications is also featured in this study. Three nuclear code packages are examined including Scale 6.1 (Scale 6.2 beta-3 for nuclear data uncertainty reporting and evaluation), Monte Carlo N-Particle (MCNP) and Parallel Environment Neutral-particle TRANsport (PENTRAN). Each of these code packages employs different neutron transport methods and cross-section evaluation. These code results are compared and contrasted for the researcher to gain perspective into if and how nuclear forensic analysis is affected by these relative outcomes from the neutronics packages. Notably, Scale 6.2 beta-3 offers perspective on the nuclear data uncertainty and how it affects final conclusions on isotopic reporting and material provenance.
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Gong, Helin. "Data assimilation with reduced basis and noisy measurement : Applications to nuclear reactor cores." Thesis, Sorbonne université, 2018. http://www.theses.fr/2018SORUS189.
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Le but de la thèse est d'améliorer l'interprétation physique et numérique de l'information impliquée dans l'assimilation de données avec des stratégies de réduction de modèles modernes et efficaces pour les systèmes gouvernés par des EDPs. Plus précisément, l'accent mis sur la tâche d'assimilation des données est lié à l'estimation de l'état pour les problèmes stationnaires, en particulier l'estimation de l'état neutronique dans les applications aux réacteurs nucléaires. Dans la première partie de la thèse, nous analysons et adaptons les approches GEIM (Generalized Empirical Interpolation Method) et PBDW (Parametrized-Background Data-Weak) du problème d'estimation d’état. Nous formulons l'analyse de stabilité pour GEIM/PBDW. Ensuite, nous proposons des approches dites « contraintes-stabilisées » (CS-GEIM / CS-PBDW) pour améliorer les performances de stabilité vis-à-vis de mesures bruitées. Une forme fermée dite GEIM / PBDW régularisée (R-GEIM / R-PBDW) est également proposée pour améliorer l'efficacité computationnelle. Dans la seconde partie, nous appliquons les techniques développées aux problèmes réels du partenaire industriel EDF, à savoir : i) la disposition de capteurs dans un cœur de réacteur nucléaire et ii) la reconstruction de champs neutroniques avec des mesures avec ou sans bruit. Les tests numériques confirment la faisabilité des techniques développées pour répondre au problème important et inévitable des mesures bruitées dans le domaine de l'assimilation de données par base réduite. Dans la troisième partie, nous fournissons des matériaux supplémentaires en i) traitant des échecs de mesure pour l'assimilation de données avec une base réduite, en particulier, EIM, comme un problème pratique; et ii) traitant de la méthode d'échantillonnage adaptatif pour fournir plus de possibilités dans les problèmes d'ingénierie avec espace des paramètres de grande dimensionThe goal of the thesis is to improve the physical and numerical interpretation of the information involved in data assimilation with modern and efficient model reduction strategies for systems held by PDEs. Specifically, the focus on the data assimilation task is related with the state estimation for stationary problems, especially neutronic state estimation in nuclear reactor applications. In the first part of the thesis, we analyze and adapt the generalized empirical interpolation method (GEIM) and the parametrized-background data-weak (PBDW) approach to the state estimation problem. We formulate the stability analysis for GEIM/PBDW. Then we propose the so-called constrained stabilized GEIM/PBDW (CS-GEIM/CS-PBDW) approaches to improve the stability performance with respect to noisy measurements. A closed form so-called regularized GEIM/PBDW (R-GEIM/R-PBDW) are also proposed to improve the computational efficiency. In the second part we apply the developed techniques to real case problems provided by the industrial partner EDF, namely, i) sensor placement in a nuclear reactor core and ii) neutronic field reconstruction with noisy or noise-free measurements. Numerical tests confirm the feasibility of developed techniques to address the important and inevitable concern of noisy measurements in the field of data assimilation with reduced basis. In the third part we provide supplementary materials in i) dealing with measurement failures for data assimilation with reduced basis, particularly, EIM, as a practical issue; and ii) dealing with the adaptive sampling method to provide more potential for engineering problems with high-dimensional parameter space
15
Faure, Bastien. "Development of neutronic calculation schemes for heterogeneous sodium-cooled nuclear cores in the Apollo3 code : application to the ASTRID prototype." Thesis, Aix-Marseille, 2019. http://www.theses.fr/2019AIXM0289.
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Les réacteurs nucléaires refroidis au sodium offrent des perspectives intéressantes pour la filière nucléaire (utilisation optimale de l'uranium naturel, réduction de la radiotoxicité des déchets nucléaires). Cependant, la nécessité d’élever le niveau de sûreté de ces réacteurs aux standards du XXIe siècle a conduit à des designs de cœurs très hétérogènes.Ainsi, les objectifs de la thèse sont l’identification des principaux phénomènes physiques devant être pris en compte lors du calcul neutronique de cœurs hétérogènes en spectre rapide, ainsi que le développement de schémas de calcul adaptés dans le code APOLLO3 du CEA. Après quelques rappels théoriques et méthodologiques, ce document présente une analyse critique des schémas de calcul disponibles dans APOLLO3 pour les réacteurs refroidis au sodium. Cette analyse permet de mettre en évidence la nécessité de simuler, dès l’étape de préparation des sections efficaces, des modes angulaires du flux qui soient représentatifs de la configuration géométrique du cœur. Pour répondre à ce besoin dans le cadre de géométries présentant une forte hétérogénéité axiale, une approximation 2D/1D à l'équation du transport des neutrons 3D est développée. Cette dernière permet de représenter de manière cohérente, et à moindre coût, des effets d’anisotropie axiale dans des calculs 2D. Une nouvelle modélisation de type traverse de l’interface cœur / réflecteur est également proposée, ainsi qu’une méthode de calcul innovante des barres de contrôle. Ces méthodes permettent, in fine, de définir un schéma de calcul de référence unique et validé numériquement, adapté à la modélisation des cœurs de réacteurs refroidis au sodiumSodium-cooled nuclear reactors offer interesting perspectives in terms of uranium resources economy and radioactive waste management. In order to meet modern safety standards, though, increasingly complex core concepts have been proposed for this technology.Hence, the first objective of this thesis is the identification of the main physical phenomena that need to be taken into account when modeling the neutronic behavior of a heterogeneous nuclear core in a fast neutron spectrum. The second objective is the development of appropriate calculation schemes in the APOLLO3 code, developed at CEA.After a brief reminder of neutronic calculation theory and methods, this document presents a critical analysis of the neutronic calculation schemes available in APOLLO3 for sodium-cooled applications. This analysis highlights the necessity to model, during the cross section preparation phase, angular modes of the neutron flux that are representative of the core geometrical configuration. To meet this need in axially heterogeneous geometries, a 2D/1D approximation to the 3D neutron transport equation is derived and implemented in APOLLO3. In particular, it is shown that this approximation allows to consistently represent axial angular modes of the flux in 2D calculation domains. Besides, a new traverse model is proposed for the core/reflector radial interface, as well as an innovative control rod calculation method. The combination of these methods allows to define a unique, and numerically validated, reference calculation scheme in APOLLO3, suitable for the calculation of a wide range of complex sodium-cooled nuclear cores
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Schramm, Marcelo. "An algorithm for multi-group two-dimensional neutron diffusion kinetics in nuclear reactor cores." reponame:Biblioteca Digital de Teses e Dissertações da UFRGS, 2016. http://hdl.handle.net/10183/142510.
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O objetivo desta tese é introduzir uma nova metodologia para a cinética bidimensional multi- grupo de difusão de nêutrons em reatores nucleares. A metodologia apresentada usa uma aproximação polinomial em um domínio homogêneo retangular com condições de contornos não homogêneas. Como ela consiste em uma série de Taylor truncada, sua estimativa de erro varia de acordo com o tamanho do retângulo. Os coeficientes são obtidos principalmente pelas suas relações com o termo independente, que _e determinado pela equação diferencial. Estas relações são obtidas apenas pelas condições de contorno, e é demonstrado serem linearmente independentes. Um esquema numérico é feito para assegurar uma rápida convergência. Estes procedimentos feitos para um retângulo homogêneo são feitos para construir soluções para problemas de autovalor e dependentes do tempo de geometria ortogonal global com parâmetros seccionalmente constantes pelo método iterativo SOR. O autovalor dominante e sua autofunção são obtidos pelo método da potência no problema de autovalor. A solução para casos dependentes do tempo usam o método de Euler modificado na variável tempo. Quatro casos-teste clássicos são considerados para ilustração.The objective of this thesis is to introduce a new methodology for two{dimensional multi{ group neutron diffusion kinetics in a reactor core. The presented methodology uses a polyno- mial approximation in a rectangular homogeneous domain with non{homogeneous boundary conditions. As it consists on a truncated Taylor series, its error estimates varies with the size of the rectangle. The coefficients are obtained mainly by their relations with the independent term, which is determined by the differential equation. These relations are obtained by the boundary conditions only, and these relations are proven linear independent. A numerical scheme is made to assure faster convergence. The procedures done for one homogeneous rectangle are used to construct the solution of global orthogonal geometry with step{wise constant parameters steady state and time dependent problems by the iterative SOR algo- rithm. The dominant eigenvalue and its eigenfunction are obtained by the power method in the eigenvalue problem. The solution for the time dependent cases uses the modi ed Euler method in the time variable. Four classic test cases are considered for illustration.
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Frieß, Friederike Renate [Verfasser], Wolfgang [Akademischer Betreuer] Liebert, and Barbara [Akademischer Betreuer] Drossel. "Neutron-Physical Simulation of Fast Nuclear Reactor Cores / Friederike Renate Frieß ; Wolfgang Liebert, Barbara Drossel." Darmstadt : Universitäts- und Landesbibliothek Darmstadt, 2017. http://d-nb.info/1138212237/34.
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MURA, LUIZ E. C. "Caracterização dos campos neutrônicos obtidos por meio de armadilhas de nêutrons no interior do núcleo do reator nuclear IPEN/MB-01." reponame:Repositório Institucional do IPEN, 2011. http://repositorio.ipen.br:8080/xmlui/handle/123456789/9999.
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Dissertação (Mestrado)
IPEN/D
Instituto de Pesquisas Energéticas e Nucleares - IPEN-CNEN/SP
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SILVEIRA, RENATO C. da. "Avaliacao da estabilidade estrutural de contencoes metalicas de centrais nucleares." reponame:Repositório Institucional do IPEN, 2000. http://repositorio.ipen.br:8080/xmlui/handle/123456789/10795.
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Dissertacao (Mestrado)
IPEN/D
Instituto de Pesquisas Energeticas e Nucleares - IPEN/CNEN-SP
20
OLIVEIRA, JOSE R. de. "Programa computacional para estudo da estrategia de controle de um reator nuclear do tipo PWR." reponame:Repositório Institucional do IPEN, 2002. http://repositorio.ipen.br:8080/xmlui/handle/123456789/11060.
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Dissertacao (Mestrado)
IPEN/D
Instituto de Pesquisas Energeticas e Nucleares - IPEN/CNEN-SP
21
Negm, Hani Hussein. "Studies on the Optimum Geometry for a Nuclear Resonance Fluorescence Detection System for Nuclear Security Applications." Kyoto University, 2014. http://hdl.handle.net/2433/193589.
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Helvenston, Edward M. (Edward March). "Analysis of in-core experiment activities for the MIT Research Reactor using the ORIGEN computer code." Thesis, Massachusetts Institute of Technology, 2006. http://hdl.handle.net/1721.1/41591.
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Thesis (S.B.)--Massachusetts Institute of Technology, Dept. of Nuclear Science and Engineering, 2006.Includes bibliographical references (leaves 73-74).
The objective of this study is to devise a method for utilizing the ORIGEN-S computer code to calculate the activation products generated in in-core experimental assemblies at the MIT Research Reactor (MITR-II). ORIGEN-S is a nuclear depletion and decay analysis code. It accounts for all types of nuclear reactions and eliminates the need for selection of the dominant reactions that will occur in a given experiment, as must be done with the existing activity calculation method. It is expected that the new approach will be easy to use, and will produce radioactivity estimations that are generally more accurate than those produced by the existing method. The ORIGEN-S method has been developed and tested for four experiments that have been or are scheduled to be irradiated in the MITR. These experiments are the Advanced Cladding Irradiation (ACI), High Temperature Irradiation Facility (HTIF), Electric Power Research Institute Electro-Chemical Potential (EPRI ECP) loop, and Annular Fuel Test Rig (AFTR). The method has also been used to perform activation analyses for ten individual elements (plus U-235 and U-238) that are commonly found in MITR in-core experiment (ICE) assemblies. The ORIGEN-S analyses for the ACI, HTIF, and EPRI ECP experiments produced results that were relatively similar to the results produced by previous analyses that utilized the current method of activation estimation. This is because the thermal neutron capture reactions, which are major contributors to the activation of these experiments, are already well accounted for in the existing method. The results of the ORIGEN-S analysis for the AFTR, which contains fissile material, were also very similar to the results of the previous analysis, despite the fact that the previous analysis accounted for changes in flux due to fissile nuclide depletion during irradiation and the current analysis did not.
It is concluded that the activation calculation method developed should be generally adequate for all experiments irradiated in the MITR core. A possible exception involves experiments containing quantities of fissile material larger than the quantities contained in the AFTR, as these experiments could produce significant changes in neutron flux levels that would render this method inadequate.
by Edward M. Helvenston.
S.B.
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Wang, Yunzhi (Yunzhi Diana). "Evaluation of the thermal-hydraulic operating limits of the HEU-LEU transition cores for the MIT Research Reactor." Thesis, Massachusetts Institute of Technology, 2009. http://hdl.handle.net/1721.1/54479.
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Thesis (S.M. and S.B.)--Massachusetts Institute of Technology, Dept. of Nuclear Science and Engineering, 2009.Cataloged from PDF version of thesis.
Includes bibliographical references (p. 93-94).
The MIT Research Reactor (MITR) is in the process of conducting a design study to convert from High Enrichment Uranium (HEU) fuel to Low Enrichment Uranium (LEU) fuel. The currently selected LEU fuel design contains 18 plates per element, compared to the existing HEU design of 15 plates per element. A transitional conversion strategy, which consists of replacing three HEU elements with fresh LEU fuel elements in each fuel cycle, is proposed. The objective of this thesis is to analyze the thermo-hydraulic safety margins and to determine the operating power limits of the MITR for each mixed core configuration. The analysis was performed using PLTEMP/ANL ver 3.5, a program that was developed for thermo-hydraulic calculations of research reactors. Two correlations were used to model the friction pressure drop and enhanced heat transfer of the finned fuel plates: the Carnavos correlation for friction factor and heat transfer, and the Wong Correlation for friction factor with a constant heat transfer enhancement factor of 1.9. With these correlations, the minimum onset of nucleate boiling (ONB) margins of the hottest fuel plates were evaluated in nine different core configurations, the HEU core, the LEU core and seven mixed cores that consist of both HEU and LEU elements. The maximum radial power peaking factors were assumed at 2.0 for HEU and 1.76 for LEU in all the analyzed core configurations. The calculated results indicate that the HEU fuel elements yielded lower ONB margins than LEU fuel elements in all mixed core configurations. In addition to full coolant channels, side channels next to the support plates that form side coolant channels were analyzed and found to be more limiting due to higher flow resistance. The maximum operating powers during the HEU to LEU transition were determined by maintaining the minimum ONB margin corresponding to the homogeneous HEU core at 6 MW. The recommended steady-state power is 5.8 MW for all transitional cores if the maximum radial peaking is adjacent to a full coolant channel and 4.9 MW if the maximum radial peaking is adjacent to a side coolant channel.
by Yunzhi (Diana) Wang.
S.M.and S.B.
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STEFANI, GIOVANNI L. de. "Sobre a técnica de Rod Drop em medidas de reatividade integral em bancos de controle e segurança de reatores nucleares." reponame:Repositório Institucional do IPEN, 2013. http://repositorio.ipen.br:8080/xmlui/handle/123456789/10210.
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Dissertação (Mestrado)
IPEN/D
Instituto de Pesquisas Energeticas e Nucleares - IPEN-CNEN/SP
25
Hamzeh, Bahmani Hamed. "Development of novel techniques for the assessment of inter-laminar resistance in transformer and reactor cores." Thesis, Cardiff University, 2014. http://orca.cf.ac.uk/69860/.
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The main aim of this project is to investigate the influence of the inter-laminar short circuit faults on the performance of magnetic cores and develop a non-destructive method to detect these kinds of defects. The eddy current path in magnetic laminations which is magnetised by time varying magnetic field was modelled by an equivalent resistor network to calculate and predict the eddy current power losses in magnetic laminations. The model was validated over a wide range of magnetisation conditions. Based on the developed model, the influence of a wide range of magnetising frequency and peak flux density on the magnetic properties of electrical steels was studied. An experimental-analytical technique was developed to separate magnetic loss components over a wide range of magnetisation. Two electrical steel laminations, Conventional Grain Oriented (CGO) and Non-Oriented (NO), were used in the experimental work of the relevant studies. 2-D FE based modelling was performed to simulate inter-laminar faults on stacks of laminations and visualise the distribution of eddy currents in the faulted laminations. The influence of inter-laminar faults on the eddy current power loss was experimentally investigated by introducing artificial short circuits of different configurations on stacks of Epstein size laminations of GO steel. A non-destructive test method was developed to detect inter-laminar fault between the laminations of the magnetic cores by means of Flux Injection Probe (FIP). A prototype model of a FIP was developed and its application to quality assessment of transformer core laminations was investigated. The research presented here can be utilised by electrical steel manufacturers and electrical machine designers to survey the effect of inter-laminar faults on the magnetic properties of magnetic cores and their quality assessment, to reduce the risk of core damage or machine failure caused by the inter-laminar faults.
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BAPTISTA, FILHO BENEDITO D. "Redes neurais para controle de sistemas de reatores nucleares." reponame:Repositório Institucional do IPEN, 1998. http://repositorio.ipen.br:8080/xmlui/handle/123456789/10723.
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Tese (Doutoramento)
IPEN/T
Instituto de Pesquisas Energeticas e Nucleares - IPEN/CNEN-SP
27
BRAGA, CLAUDIA C. "Analise de sensibilidade para modelagem semi-mecanistica de acidentes severos." reponame:Repositório Institucional do IPEN, 1994. http://repositorio.ipen.br:8080/xmlui/handle/123456789/10399.
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Dissertacao (Mestrado)
IPEN/D
Instituto de Pesquisas Energeticas e Nucleares - IPEN/CNEN-SP
28
MUNIZ, RAFAEL O. R. "Análise neutrônica e especificação técnica para o combustível a dispersão UMo-Al com adição de veneno queimável." reponame:Repositório Institucional do IPEN, 2015. http://repositorio.ipen.br:8080/xmlui/handle/123456789/25671.
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Tese (Doutorado em Tecnologia Nuclear)
IPEN/T
Instituto de Pesquisas Energeticas e Nucleares - IPEN-CNEN/SP
29
CARNEIRO, ALVARO L. G. "Medida de distribuicao da densidade de potencia relativa do nucleo do reator IPEN/MB-01...vareta combustivel." reponame:Repositório Institucional do IPEN, 1996. http://repositorio.ipen.br:8080/xmlui/handle/123456789/10669.
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Dissertacao (Mestrado)
IPEN/D
Instituto de Pesquisas Energeticas e Nucleares - IPEN/CNEN-SP
30
Rolfo, Stefano. "LES and Hybrid RANS/LES turbulence modelling in unstructured finite volume code and applications to nuclear reactor fuel bundle." Thesis, University of Manchester, 2010. https://www.research.manchester.ac.uk/portal/en/theses/les-and-hybrid-ransles-turbulence-modelling-in-unstructured-finite-volume-code-and-applications-to-nuclear-reactor-fuel-bundle(14e99c49-c1f5-442d-926e-2324a9701690).html.
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Rod bundle is a typical constitutive element of a very wide range of nuclear reactor designs. This thesis describes the investigation of such geometry with wall-resolved Large Eddy Simulation (LES). In order to alleviate the mesh constraint, imposed by the near wall resolution, the usage of embedded refinements and polyhedral meshes is analysed firstly with a inviscid laminar case (Taylor Green vortices) and secondly with a fully turbulent case (channel flow only with embedded refinement). The inviscid test case shows that the addition of embedded refinements decreases the conservation properties of the code. Indeed the accuracy decreases from second order in a structured conformal mesh, to something in between first and second order depending on the quality of the unstructured mesh. Better results are obtained when the interface between refined and coarse areas presents a more regular and structured pattern, reducing the generation of skewed and stretched cells. The channel flow simulation shows that the Reynolds stresses, of some embedded refined meshes, are affected by spurious oscillations. Surprisingly this effect is present in the unstructured meshes with the best orthogonal properties. Indeed analysis of Reynolds stress budgets shows that terms, where the gradient in the wall normal direction is dominant, have a largely oscillatory behaviour. The cause of the problem is attributed to the convective term and in particular in the method used for the gradient reconstruction. As a consequence of these contradictory signs between the inviscid and the fully turbulent cases, the rod bundle test case is analysed using a conventional body fitted multiblock mesh. Two different Reynolds numbers are investigated reporting Reynolds stresses and budgets. The flow is characterised by an energetic and almost periodic azimuthal flow pulsation in the gap region between adjacent sub-channels, which makes turbulent quantities largely different from those in plane channel and pipes and enhances mixing. Experiments found that a constant Strouhal number, with the variation of the Reynolds number, characterises the phenomenon. The frequency analysis finds that present simulations are distinguished by three dominant frequencies, the first in agreement with the experimental value and two higher ones, which might be due to the correlation of the azimuthal velocity in the streamwise direction. Several passive temperature fields are added at the simulations in order to study the effects of the variation of the Prandtl number and the change in boundary conditions (Neumann and Dirichlet). A simplified case where an imbalance of the scalar between adjacent sub-channels is also investigated in order to evaluate the variation of the heat fluxes with respect to the homogeneous case. An alternative solution, to reduce the mesh constraint imposed by the wall, is to hybridize LES with RANS. The main achievement of this work is to integrate the heat transfer modelling to the already existing model for the dynamic part. Further investigations of the blending function, used to merge the two velocity fields, are carried out in conjunction with a study of the model dependency on the mesh resolution. The validation is performed on a fully developed channel flow at different Reynolds numbers and with constant wall heat flux. On coarse meshes the model shows an improvement of the results for both thermal and hydraulic parts with respect to a standard LES. On refined meshes, suitable for wall-resolved LES, the model suffers from a problem of double counting of modelled Reynolds stresses and heat fluxes because the RANS contribution does not naturally disappear as the mesh resolution increases.
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DOMINGOS, DOUGLAS B. "Calculos neutronicos, termo-hidrulicos e de seguranca de um dispositivo para irradiacao de miniplacas (DIM) de elementos combustiveis tipo dispersao." reponame:Repositório Institucional do IPEN, 2010. http://repositorio.ipen.br:8080/xmlui/handle/123456789/9510.
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Fundação de Amparo à Pesquisa do Estado de São Paulo (FAPESP)
Dissertacao (Mestrado)
IPEN/D
Instituto de Pesquisas Energeticas e Nucleares - IPEN-CNEN/SP
FAPESP:08/55686-6
32
CARLUCCIO, THIAGO. "Implementação e qualificação de metodologia de cálculos neutrônicos em reatores subcríticos acionados por fonte externa de nêutrons e aplicações." reponame:Repositório Institucional do IPEN, 2011. http://repositorio.ipen.br:8080/xmlui/handle/123456789/10033.
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Tese (Doutoramento)
IPEN/T
Instituto de Pesquisas Energeticas e Nucleares - IPEN-CNEN/SP
33
Connolly, Kevin John. "A coarse mesh radiation transport method for reactor analysis in three dimensional hexagonal geometry." Diss., Georgia Institute of Technology, 2012. http://hdl.handle.net/1853/50149.
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A new whole-core transport method is described for 3-D hexagonal geometry. This is an extension of a stochastic-deterministic hybrid method which has previously been shown highly accurate and efficient for eigenvalue problems. Via Monte Carlo, it determines the solution to the transport equation in sub-regions of reactor cores, such as individual fuel elements or sections thereof, and uses those solutions to compose a library of response expansion coefficients. The information acquired allows the deterministic solution procedure to arrive at the whole core solution for the eigenvalue and the explicit fuel pin fission density distribution more quickly than other transport methods. Because it solves the transport equation stochastically, complicated geometry may be modeled exactly and therefore heterogeneity even at the most detailed level does not challenge the method. In this dissertation, the method is evaluated using comparisons with full core Monte Carlo reference solutions of benchmark problems based on gas-cooled, graphite-moderated reactor core designs. Solutions are given for core eigenvalue problems, the calculation of fuel pin fission densities throughout the core, and the determination of incremental control rod worth. Using a single processor, results are found in minutes for small cores, and in no more than a few hours for a realistically large core. Typical eigenvalues calculated by the method differ from reference solutions by less than 0.1%, and pin fission density calculations have average accuracy of well within 1%, even for unrealistically challenging core configuration problems. This new method enables the accurate determination of core eigenvalues and flux shapes in hexagonal cores with efficiency far exceeding that of other transport methods.
34
Vaidya, Udyanth. "Uncertainty & Sensitivity Analysis of Nuclear Fuel Using Transuranus & Dakota." Thesis, KTH, Fysik, 2021. http://urn.kb.se/resolve?urn=urn:nbn:se:kth:diva-302565.
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With the initiative taken by the SUNRISE project (Sustainable Nuclear Energy Research in Sweden) to construct a Lead-cooled research reactor, this thesis intends to extend the knowledge within nuclear fuel development. By using integral iterative modelling and simulating techniques that mimic real-world phenomena, novel fuel materials like uranium nitride are assessed for future validation. The work deals with the fuel performance analysis of the SUNRISE LFR, employing the TRANSURANUS fuel performance code. This code contains a collection of model parameters that simulate the thermo-mechanical behaviour of the fuel cladding system on an engineering scale of the reactor core. A comparative study is performed for UO$_2$ and UN fuels using the same input data such as fuel geometry. In addition, predefined information relating to the neutronics analysis for the reactor was used as input to the TRANSURANUS code along with literature reviews to select the accurate models on the reactor, fuel, and its behaviour. Furthermore, a sensitivity study is carried out to assess the models and parameters affected by more significant uncertainty. The uncertainty analysis of the UN fuel's swelling models is performed using the Dakota toolkit. The sampling of input data using the Dakota software coupled with the nuclear simulation program TRANSURANUS produced partial rank correlation coefficients significant to the modelling. However, since the assessed models displayed the same correlation coefficients, the results conclude that a deeper understanding of the theoretical swelling model might be required.I samverkan med initiativet av SUNRISEprojektet (Sustainable Nuclear Energy Research inSweden) som syftar att bygga en blykyld forskningsreaktor, avser denna avhandling att utökakunskapen inom kärnbränsleutveckling. Med användning av integral iterativ modellering ochsimuleringstekniker som efterliknar verkliga fenomen bedöms nya bränslematerial somuranmononitrid för framtida validering. Arbetet behandlar analysen av bränsleprestanda för SUNRISE LFR, med användning avTRANSURANUS bränsleprestandakod. Denna kod innehåller en samling modellparametrarsom simulerar det termomekaniska beteendet hos bränslebetäckningssystemet i en tekniskskala för reaktorkärnan. En jämförande studie utförs för UO2 och UN-bränslen med sammaingångsdata som t.ex bränslegeometrin. Dessutom användes fördefinierad information om denneutroniska analysen för reaktorn som ingångsdata till TRANSURANUSkoden tillsammans medgranskning av litteratur för att välja lämpliga modeller för reaktorn, bränslet och dess beteende.Därtill genomfördes en känslighetsstudie för att bedöma de modeller och parametrar sompåverkas av mer betydande osäkerhet. Osäkerhetsanalysen av UN-bränslets svällningsmodeller utförs med hjälp av Dakota-verktyget.Samlingen av indata med Dakota-programmet i kombination medkärnkraftssimuleringsprogrammet TRANSURANUS gav korrelationskoefficienter för partiell rangviktiga för modelleringen. Eftersom de utvärderade modellerna visade sammakorrelationskoefficienter, tyder slutsatsen på att en djupare förståelse av den teoretiskasvällningsmodellen krävs
35
ROSSI, LUBIANKA F. R. "Acoplamento entre os métodos diferencial e da teoria da perturbação para o cálculo dos coeficientes de sensibilidade em problemas de transmutação nuclear." reponame:Repositório Institucional do IPEN, 2014. http://repositorio.ipen.br:8080/xmlui/handle/123456789/23594.
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Tese (Doutorado em Tecnologia Nuclear)
IPEN/T
Instituto de Pesquisas Energeticas e Nucleares - IPEN-CNEN/SP
36
REIS, REGIS. "Análise do comportamento sob irradiação do combustível nuclear a altas queimas com os programas computacionais FRAPCON e FRAPTRAN." reponame:Repositório Institucional do IPEN, 2014. http://repositorio.ipen.br:8080/xmlui/handle/123456789/11797.
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Dissertação (Mestrado em Tecnologia Nuclear)
IPEN/D
Instituto de Pesquisas Energeticas e Nucleares - IPEN-CNEN/SP
37
Leduc, Christian. "Modélisation de la condensation en film sur les parois d'une enceinte de réacteurs." Université Joseph Fourier (Grenoble ; 1971-2015), 1995. http://www.theses.fr/1995GRE10157.
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Un code enceinte, utilise dans les analyses de surete des centrales nucleaires, doit pouvoir predire les evolutions de la pression et des concentrations en vapeur, air et hydrogene dans l'enceinte de confinement d'un reacteur a eau pressurisee en situation accidentelle. La condensation de la vapeur sur les parois froides est un facteur essentiel de ces evolutions. Nous proposons un modele de condensation en film en presence de gaz incondensables. L'ecoulement du film est considere laminaire. Trois approches differentes sont abordees pour modeliser les transferts dans les couches limites: utilisation de correlations globales dans lesquelles nous avons employe un nombre de grashof hybride exprimant que la convection est a la fois de nature thermique et massique. Calcul de la couche limite avec utilisation de lois de paroi pour un regime de convection forcee. Nous proposons des lois prenant en compte la vitesse d'aspiration de la couche limite due a la condensation, et l'influence de l'etat de surface du film (lisse ou ridee). Calcul de la couche limite avec un modele k-epsilon a bas nombre de reynolds pour un regime de convection naturelle. Ces modeles ont ete implantes dans le code de thermohydraulique 3d-trio-vf. Les coefficients de transfert de chaleur theoriques obtenus sont compares a ceux des resultats experimentaux
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MURA, LUIS F. L. "Determinação experimental de taxas de reação no 238U e 235U ao longo do raio da pastilha de UO2 do reator IPEN/MB-01." reponame:Repositório Institucional do IPEN, 2015. http://repositorio.ipen.br:8080/xmlui/handle/123456789/25670.
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Tese (Doutorado em Tecnologia Nuclear)
IPEN/T
Instituto de Pesquisas Energeticas e Nucleares - IPEN-CNEN/SP
39
CARNEIRO, ALVARO L. G. "Desenvolvimento de sistema de monitoracao e diagnostico aplicado a valvulas moto-operadas utilizadas em centrais nucleares." reponame:Repositório Institucional do IPEN, 2003. http://repositorio.ipen.br:8080/xmlui/handle/123456789/11109.
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Tese (Doutoramento)
IPEN/T
Instituto de Pesquisas Energeticas e Nucleares - IPEN/CNEN-SP
40
NUNES, BEATRIZ G. "Determinação exerimental de razões espectrais e do espectro de energia dos nêutrons no combustível do reator nuclear IPEN/MB-01." reponame:Repositório Institucional do IPEN, 2012. http://repositorio.ipen.br:8080/xmlui/handle/123456789/10069.
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Dissertação (Mestrado)
IPEN/D
Instituto de Pesquisas Energeticas e Nucleares - IPEN-CNEN/SP
41
PERROTTA, JOSE A. "Proposta de um nucleo de reator PWR avancado com caracteristicas adequadas para o conceito de seguranca passiva." reponame:Repositório Institucional do IPEN, 1999. http://repositorio.ipen.br:8080/xmlui/handle/123456789/10704.
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Tese (Doutoramento)
IPEN/T
Instituto de Pesquisas Energeticas e Nucleares - IPEN/CNEN-SP
42
SILVESTRE, LARISSA J. B. "PCRELAP5 - Programa de cálculo para os dados de entrada do código RELAP5." reponame:Repositório Institucional do IPEN, 2016. http://repositorio.ipen.br:8080/xmlui/handle/123456789/26393.
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Dissertação (Mestrado em Tecnologia Nuclear)
IPEN/D
Instituto de Pesquisas Energeticas e Nucleares - IPEN-CNEN/SP
43
Trinh, Ngoc Duy. "Emission de neutrons par les réactions d'ions lourds (4,6-95 MeV/nucléon)." Thesis, Normandie, 2018. http://www.theses.fr/2018NORMC234/document.
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Les accélérateurs d’ions lourds sont un outil incontournable pour la recherche en physique nucléaire. Ils sont également utilisés pour diverses applications. Il est nécessaire de caractériser la production des neutrons secondaires dans les accélérateurs afin de garantir un fonctionnement sûr en toutes circonstances. Cependant, les données expérimentales sont très rares voire inexistantes. Pour certaines données, on note des divergences entre différentes publications. Des désaccords sont aussi observés entre les mesures et les calculs. Toutes ces raisons justifient le programme Thick Target Neutron Yields (TTNY) dont l’objectif est de mesurer des spectres doublement différentiels (énergie, angle) des neutrons générés par l’interaction des ions lourds (12≤Afaisceau≤208 et 4,6 MeV/nucléon≤Efaisceau≤95 MeV/nucléon) sur cibles épaisses (natC, natCu et natNb). Deux techniques de mesure ont été utilisées : Activation et Temps de vol. Cela permet d’avoir une meilleure confiance dans les mesures, d’étudier les limites expérimentales et de consolider les conclusions que l’on peut en tirer. Les mesures sont comparées à des simulations effectuées dans ce travail avec les codes Monte-Carlo les plus utilisés en calcul nucléaires : PHITS (japonais), FLUKA (européen (CERN/INFN)) et MCNP (américain). Ces comparaisons ont permis d’évaluer la qualité des codes dans les énergies étudiées et pour les masses des noyaux explorées. Elles ont permis aussi de conclure sur les incertitudes systématiques et les éventuelles évolutions à apporter aux modèles physiques de ces codesHeavy-ion accelerators are an essential tool for nuclear physics research. They are also adopted in several applications. It is necessary to characterize the secondary neutrons production in order to guarantee a safe operation in every circumstance in accelerators. However, experimental data are very rare or even non-existent. For some data, we notice disagreements between different publications. Disagreements are also observed between measurements data and simulations. For all these reasons, we established the program Thick Target Neutron Yields (TTNY). This program aims to measure the double differential neutron spectra (energy, angle) generated by the interactions of heavy-ions (12≤Abeam≤208 and 4.6 MeV/nucleon≤Ebeam≤95 MeV/nucleon) on thick targets (natC, natCu and natNb). Two measurements methods were adopted: Activation and Time of Flight. This choice allows having a better confidence on the measurements, studying experimental limits and consolidating the conclusions that could be drawn from the experimental results. The measurements are compared to the simulations performed with some Monte-Carlo widely used in nuclear simulation: PHITS (Japanese), FLUKA (European (CERN/INFN)) and MCNP (American). These comparisons allowed evaluating the modeling quality of heavy-ion reactions for the energies and masses explored in this work. We also conclude on the systematic uncertainties and on the potential improvements to be introduced to physics models of these codes
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PERRENOUD, HELENA G. "Modulo de extracao de eventos em assinaturas de potencia de valvulas moto-operadas, usando um sistema especialista para o sistema de diagnostico de MOV's utilizado em reatores nucleares." reponame:Repositório Institucional do IPEN, 2001. http://repositorio.ipen.br:8080/xmlui/handle/123456789/10967.
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Dissertacao (Mestrado)
IPEN/D
Instituto de Pesquisas Energeticas e Nucleares - IPEN/CNEN-SP
45
ALBUQUERQUE, LEVI B. de. "Categorizacao de tensoes em modelos de elementos finitos de conexoes bocal-vaso de pressao." reponame:Repositório Institucional do IPEN, 1999. http://repositorio.ipen.br:8080/xmlui/handle/123456789/10761.
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Dissertacao (Mestrado)
IPEN/D
Instituto de Pesquisas Energeticas e Nucleares - IPEN/CNEN-SP
46
CRUZ, JULIO R. B. "Procedimento analitico para previsao do comportamento estrutural de componentes truncados." reponame:Repositório Institucional do IPEN, 1998. http://repositorio.ipen.br:8080/xmlui/handle/123456789/10665.
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Tese(Doutoramento)
IPEN/T
Instituto de Pesquisas Energeticas e Nucleares - IPEN/CNEN-SP
47
HIRATA, DANIEL M. "Estimativa da frequencia de danos ao nucleo devido a perda de refrigerante primario e bloqueio de canal de refrigeracao do reator de pesquisas IEA-R1 do IPEN-CNEN/SP - APS nivel 1." reponame:Repositório Institucional do IPEN, 2009. http://repositorio.ipen.br:8080/xmlui/handle/123456789/9483.
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Dissertacao (Mestrado)
IPEN/D
Instituto de Pesquisas Energeticas e Nucleares - IPEN-CNEN/SP
48
Gerardin, Delphine. "Développement de méthodes et d’outils numériques pour l’étude de la sûreté du réacteur à sels fondus MSFR." Thesis, Université Grenoble Alpes (ComUE), 2018. http://www.theses.fr/2018GREAI068/document.
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Les travaux réalisés pendant cette thèse portent sur l’étude de la sûreté du Molten Salt Fast Reactor (MSFR) et incluent à la fois des méthodes d’analyse de risques et des calculs déterministes de sûreté et de design. Ce travail s’inscrit dans le cadre du projet européen SAMOFAR.Le MSFR est un réacteur régénérateur à spectre neutronique rapide qui fonctionne en cycle thorium dans sa configuration de référence, établie en début du projet SAMOFAR. Il a été sélectionné par le Forum International Génération IV pour son potentiel prometteur. Comme tout réacteur nucléaire de quatrième génération, il doit répondre à différentes contraintes dont une sûreté optimale. Celle-ci doit être étudiée dès le stade de conception afin d’être intégrée au design lors de sa définition plutôt qu’ajoutée a posteriori. En raison de ses spécificités, en particulier l’état liquide du combustible, et du stade préliminaire de son design, l’analyse de sûreté du MSFR nécessite l’utilisation de méthodologies d’analyse de sûreté adaptées et technologiquement neutres. Dans cette thèse, une telle méthodologie a été développée et une première application au MSFR réalisée. Elle a notamment permis d’identifier les évènements initiateurs d’accident de ce réacteur et d’élaborer une liste resserrée d’évènements à traiter dans la suite de l’analyse de sûreté.D’autre part, un nouveau code système a été développé pour les études de sûreté. Il est basé sur la diffusion neutronique, prend en compte le transport des précurseurs de neutrons retardés et la puissance résiduelle du combustible. Il a été utilisé pour simuler les transitoires associés à certains des évènements initiateurs et évaluer leurs conséquences pour définir, par la suite, des systèmes de protection adaptés. Ce travail a confirmé l’importance d’un dispositif spécifique au MSFR, le système de vidange d’urgence, permettant de vidanger le combustible en cas d’accident en cœur. Des études paramétriques ont été menées afin de dimensionner ce système avec pour objectif d'assurer l’évacuation de la chaleur résiduelle du combustible et sa sous-criticité en toutes circonstances.Enfin, une première ébauche de l’architecture de sûreté du réacteur a été proposée incluant l’identification des systèmes de protection et la définition des barrières de confinement. Les études de sûreté ont permis de faire des retours sur le design initialement défini. Ils incluent l’ajout de composants, des propositions de design alternatifs, et soulignent les manques de connaissances sur certains phénomènes ou procédures. L’analyse de sûreté réalisée remplit ainsi son objectif principal : guider le design du réacteur dès sa conception afin d’en améliorer la sûretéThis PhD thesis focuses on the study of the Molten Salt Fast Reactor (MSFR) safety. It includes risk analysis methods and deterministic computations for the safety and the design of the reactor. This work was performed in the frame of the SAMOFAR European project.The MSFR is an is-breeder reactor with a fast neutron spectrum. In its reference configuration, defined at the beginning of the SAMOFAR project, it works with the thorium fuel cycle. The MSFR was selected by the Generation IV international forum for its promising features. As any fourth-generation reactor, it must fulfill several objectives including an improved safety. Thus, safety studies should be performed from the early design phases to achieve a safety that is built-in the design rather than added-on. Because of the unique characteristics of the MSFR, including a liquid circulating fuel, and its preliminary design phase, the safety assessment of the reactor should rely on adapted and technological neutral methodologies. In this PhD, such a methodology was developed and a first application to the MSFR was carried on. It allowed to identify the initiating events of the reactor and to elaborate a restricted list of events to be studied in the next steps of the safety analysis.Furthermore, a new code system was developed for the safety studies. It is based on neutronic diffusion and takes into account the movement of the delayed neutrons precursors and the production of the residual heat in the fuel. It was used to simulate the transients associated to some of the identified initiating events with the objective to evaluate their consequences and the need for adequate protection systems. This work confirmed the importance of a device that is specific to the MSFR: the emergency draining system (EDS). It allows to drain the fuel in case of accident in the core. Parametric studies were then carried on for the sizing of the EDS with the objective to ensure the evacuation of the residual heat and the sub-criticality of the system under any circumstances.Finally, a first version of the safety architecture was proposed with the identification of the protection systems and the definition of the confinement barriers. Thanks to the safety studies, feedbacks on the initial design were made to enhance the safety the reactor. They include the addition of new components, the modification of some systems and they highlight the lack of knowledge on some phenomena or procedure. In that respect, the safety analysis fulfil its main objective: to influence the design of the reactor since its conception in order to improve its safety
49
CASTRO, ALFREDO J. A. de. "Análise experimental de velocidade crítica em elemento combustível tipo placa plana para reatores nucleares de pesquisa." reponame:Repositório Institucional do IPEN, 2017. http://repositorio.ipen.br:8080/xmlui/handle/123456789/28022.
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Conselho Nacional de Desenvolvimento Científico e Tecnológico (CNPq)
Os elementos de combustível de um reator nuclear de pesquisa tipo MTR (\"Material Testing Reactor\") são, em sua grande maioria, formados por placas de combustível revestidas com alumínio contendo no cerne silicileto de urânio (U3Si2) disperso em matriz de alumínio. Essas placas possuem espessura da ordem de milímetros e comprimentos muito maiores em relação à sua espessura. Elas são dispostas paralelamente no conjunto que forma o elemento combustível, de maneira a formar canais entre elas com poucos milímetros de espessura, por onde escoa o fluido de refrigeração (água leve ou água pesada). Essa configuração, associada à necessidade de um escoamento com altas vazões para garantir o resfriamento das placas em operação, pode gerar problemas de falhas mecânicas das placas de combustível devido às vibrações induzidas pelo escoamento nos canais e, consequentemente, acidentes de proporções graves no caso de velocidade crítica que possa gerar o colapso das placas. Embora não haja ruptura das placas de combustível durante o colapso, as deflexões permanentes excessivas das placas podem causar bloqueio do canal de escoamento no núcleo do reator e levar ao superaquecimento nas placas. Para este trabalho, foram desenvolvidas uma bancada experimental com capacidade para altas vazões volumétricas (Q=100 m3/h) e uma seção de testes que simula um elemento combustível do tipo placa com três canais de resfriamento. A seção de testes foi construída com placas de alumínio e acrílico e foi instrumentada com sensores de deformação, sensores de pressão, um acelerômetro e um tubo de pitot. As dimensões da seção de testes foram baseadas nas dimensões do Elemento Combustível do Reator Multipropósito Brasileiro (RMB), cujo projeto está sendo coordenado pela Comissão Nacional de Energia Nuclear - CNEN. Os experimentos realizados alcançaram o objetivo de chegar à condição de velocidade crítica de Miller com o colapso das placas. A velocidade crítica foi atingida com 14,5 m/s levando a consequente deformação plástica das placas que formam o canal do escoamento. O canal central na entrada da seção de testes apresentou uma abertura de 3 mm em seu centro, causando um grande bloqueio do escoamento nos canais laterais. Este comportamento foi v constatado visualmente durante a desmontagem da seção de testes, ilustrado e discutido na análise de resultados apresentado neste trabalho. O bloqueio dos canais também foi observado por meio de gráficos de queda de pressão e por gráficos das deformações da entrada, centro e saída das placas contra a velocidade média da seção de testes. Observou-se uma queda da resistência hidráulica da seção de testes devido ao aumento da seção transversal de escoamento no canal central e um aumento exponencial das deformações quando da ocorrência da velocidade crítica. Comparativamente, o valor experimental obtido para velocidade crítica na seção de testes foi da ordem de 85% do valor obtido por cálculo com a expressão teórica de Miller. Os experimentos realizados permitiram um melhor entendimento da interação fluido estrutura em elementos de combustível tipo placa como: valores de frequências de vibrações naturais, instabilidade fluido elástica e desenvolvimento de técnicas para a detecção de valores de velocidade crítica.
Tese (Doutorado em Tecnologia Nuclear)
IPEN/T
Instituto de Pesquisas Energéticas e Nucleares - IPEN-CNEN/SP
CNPq:481193/2012-0
50
Benoit, Jean-christophe. "Développement d’un code de propagation des incertitudes des données nucléaires sur la puissance résiduelle dans les réacteurs à neutrons rapides." Thesis, Paris 11, 2012. http://www.theses.fr/2012PA112254/document.
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Ce travail de thèse s’inscrit dans le domaine de l’énergie nucléaire, de l’aval du cycle du combustible et du calcul des incertitudes. Le CEA doit concevoir le prototype ASTRID, réacteur à neutrons rapides refroidi au sodium (RNR), qui est l’un des concepts retenus au sein du forum Génération IV et dont la puissance résiduelle et l’estimation de son incertitude ont un impact important. Ce travail consiste à développer un code de propagation des incertitudes des données nucléaires sur la puissance résiduelle dans les RNR.La démarche s’est déroulée en trois temps.La première étape a permis de limiter le nombre de paramètres intervenant dans le calcul de la puissance résiduelle. Pour cela, un essai de puissance résiduelle sur le réacteur PHENIX (PUIREX 2008) a été interprété de façon à valider expérimentalement le formulaire d’évolution DARWIN pour les RNR et à quantifier les termes sources de la puissance résiduelle.La deuxième étape a eu pour but de développer un code de propagation des incertitudes : CyRUS (Cycle Reactor Uncertainty and Sensitivity). Une méthode de propagation déterministe a été retenue car elle permet des calculs rapides et fiables. Les hypothèses de linéarité et de normalité qu’elle entraîne ont été validées théoriquement. Le code a également été comparé avec succès à un code stochastique sur l’exemple de la fission élémentaire thermique de l’235U.La dernière partie a été une application du code sur des expériences de puissance résiduelle d’un réacteur, de bilan matière d’une aiguille combustible et d’une fission élémentaire de l’235U. Le code a démontré des possibilités de retour d’expériences sur les données nucléaires impactant l’incertitude de cette problématique.Deux résultats principaux ont été mis en évidence. Tout d’abord, les hypothèses simplificatrices des codes déterministes sont compatibles avec un calcul précis de l’incertitude de la puissance résiduelle. Ensuite, la méthode développée est intrusive et permet un retour d’expérience sur les données nucléaires des expériences du cycle. En particulier, ce travail a montré qu’il est déterminant de mesurer précisément les rendements de fission indépendants et de déterminer leurs matrices de covariances afin d’améliorer la précision du calcul de la puissance résiduelleThis PhD study is in the field of nuclear energy, the back end of nuclear fuel cycle and uncertainty calculations. The CEA must design the prototype ASTRID, a sodium cooled fast reactor (SFR) and one of the selected concepts of the Generation IV forum, for which the calculation of the value and the uncertainty of the decay heat have a significant impact. In this study is developed a code of propagation of uncertainties of nuclear data on the decay heat in SFR.The process took place in three stages.The first step has limited the number of parameters involved in the calculation of the decay heat. For this, an experiment on decay heat on the reactor PHENIX (PUIREX 2008) was studied to validate experimentally the DARWIN package for SFR and quantify the source terms of the decay heat.The second step was aimed to develop a code of propagation of uncertainties : CyRUS (Cycle Reactor Uncertainty and Sensitivity). A deterministic propagation method was chosen because calculations are fast and reliable. Assumptions of linearity and normality have been validated theoretically. The code has also been successfully compared with a stochastic code on the example of the thermal burst fission curve of 235U.The last part was an application of the code on several experiments : decay heat of a reactor, isotopic composition of a fuel pin and the burst fission curve of 235U. The code has demonstrated the possibility of feedback on nuclear data impacting the uncertainty of this problem.Two main results were highlighted. Firstly, the simplifying assumptions of deterministic codes are compatible with a precise calculation of the uncertainty of the decay heat. Secondly, the developed method is intrusive and allows feedback on nuclear data from experiments on the back end of nuclear fuel cycle. In particular, this study showed how important it is to measure precisely independent fission yields along with their covariance matrices in order to improve the accuracy of the calculation of the decay heat