Dissertations / Theses on the topic 'Nuclear fuels – Computer programs'

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Dissertação (Mestrado)
IPEN/D
Instituto de Pesquisas Energeticas e Nucleares - IPEN-CNEN/SP

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Dissertacao (Mestrado)
IPEN/D
Instituto de Pesquisas Energeticas e Nucleares - IPEN/CNEN-SP

O objetivo deste trabalho é verificar os resultados fornecidos pelo programa computacional FRAPCON-3, hoje na versão 5, utilizado no processo de simulação do comportamento de varetas combustíveis de reatores refrigerados a água pressurizada (Pressurized Water Reactor PWR), sob situações operacionais de regime permanente, em condições de alta queima. Para realizar a verificação, foi utilizada a base de dados FUMEX-III, que fornece dados relativos a experimentos realizados com diversos tipos de combustíveis nucleares, submetidos a diversas condições operacionais. Através dos resultados obtidos na simulação do programa FRAPCON-3.5 e da sua comparação com os dados experimentais da base FUMEX-III, foi possível constar que o programa possui boa capacidade de predizer o comportamento operacional da vareta combustível em regime permanente a altas queimas. O trabalho consiste também em verificar a correlação entre UO2 e UO2-7%Gd2O3 na análise dos modelos que simulam o comportamento das pastilhas combustível. A adição do óxido de gadolínio ou gadolínia (Gd2O3), constitui-se na opção tecnológica mais solidamente consagrada e hoje comum em várias centrais nucleares. Por meio dos resultados obtidos nas simulações computacionais foram apresentadas e discutidas a influência das propriedades do UO2 e UO2-7%Gd2O3, quanto à temperatura no centro do combustível, liberação de gás de fissão na vareta, temperatura média do revestimento, volume interno e pressão interna da vareta combustível.
The objective of this work is to verify the results provided by the computer program FRAPCON-3, now in version 5, used in the simulation process of the behavior of fuel rods of pressurized water reactors - PWR permanent, in conditions of high burn. In order to carry out the verification, the FUMEX-III database was used, which provides data on experiments performed with different types of nuclear fuel, under different operating conditions. The results obtained in the simulation of the FRAPCON-3.5 program and its comparison with the experimental data of the FUMEX-III base showed that the program has a good ability to predict the operational behavior of the fuel rod in a steady state at high burn. The work also consists in verifying the correlation between UO2 and UO2-7%Gd2O3 in the analysis of models that simulate the behavior of fuel pellets. The addition of gadolinium oxide (Gd2O3) constitutes the most solidly established and now common technological option in several nuclear power plants. The influence of the properties of UO2 and UO2-7%Gd2O3 on the temperature at the center of the fuel, fission gas release on the rod, average coating temperature, internal volume and pressure were presented and discussed. of the fuel rod.

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Dissertacao (Mestrado)
IPEN/D
Instituto de Pesquisas Energeticas e Nucleares - IPEN/CNEN-SP

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Tese (Doutorado em Tecnologia Nuclear)
IPEN/T
Instituto de Pesquisas Energeticas e Nucleares - IPEN-CNEN/SP

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

Nuclear and gas turbine power plants, computer chips, and other devices and industries are running hotter than ever for longer than ever. With no apparent end to the trend, the potential arises for a phenomenon known as thermotransport to cause undesirable changes in these high temperature materials. The diffuse-interface method known as the phase-field model is a useful tool in the simulation and prediction of thermotransport driven microstructure evolution in materials. The objective of this work is to develop a phase-field model using practical and empirical properties of thermodynamics and kinetics for simulating the interdiffusion behavior and microstructural evolution of single and multiphase binary alloy system under composition and/or temperature gradients. Simulations are carried out using thermodynamics and kinetics of real systems, such as the U-Zr solid metallic fuel, with emphasis on the temperature dependencies of the kinetics governing diffusional interactions in single-phase systems and microstructural evolution in the presence of multiple driving forces in multi-phase systems. A phase field model is developed describing thermotransport in the γ phase of the U-Zr alloy, a candidate for advanced metallic nuclear fuels. The model is derived using thermodynamics extracted from the CALPHAD database and temperature dependent kinetic parameters associated with thermotransport from the literature. Emphasis is placed upon the importance of the heat of transport, Q*, and atomic mobility, β. Temperature dependencies of each term are estimated from empirical data obtained directly from the literature, coupled with the textbook phenomenological formulae of each parameter. A solution is obtained via a finite volume approach with the aid of the FiPy® partial differential equation solver. Results of the simulations are described based on individual flux contributions from the gradients of both composition and temperature, and are found to be remarkably similar to experimental results from the literature. In an additional effort the thermotransport behavior of a binary two-phase alloy is modeled, for the first time, via the phase-field method for a two-phase (γ + β) U-Zr system. The model is similarly built upon CALPHAD thermodynamics describing the γ and β phases of the U-Zr system and thermotransport parameters for the γ phase from literature. A parametric investigation of how the heats of transport for U and Zr in the β phase affect the redistribution is performed, and the interplay between system kinetics and thermodynamics are examined. Importantly, a strict control over the microstructure that is placed into the temperature gradient (at t=0) is used to eliminate the randomness associated with microstructural evolution from an initially unstable state, allowing an examination of exactly how the β phase thermotransport parameters affect the redistribution behavior of the system. Results are compared to a control scenario in which the system evolves only in the presence of thermodynamic driving forces, and the kinetic parameters that are associated with thermotransport are negligible. In contrast to the single-phase simulations, in the presence of a large thermodynamic drive for phase transformation and stability, the constituent redistribution caused by the thermotransport effect is comparatively smaller.
ID: 031001396; System requirements: World Wide Web browser and PDF reader.; Mode of access: World Wide Web.; Title from PDF title page (viewed June 3, 2013).; Thesis (M.S.M.S.E.)--University of Central Florida, 2012.; Includes bibliographical references (p. 50-53).
M.S.M.S.E.
Masters
Materials Science Engineering
Engineering and Computer Science
Materials Science and Engineering

Highly stressful situations have a prominent influence on the human psyche and it is crucial to take this into account when designing and evaluating software used predominantly in nerve wrecking situations. Lack of principles for design and evaluation of design of emergency handling software programs allows for graphic interface solutions that leave room for mistakes and slips in an environment where making an error can have catastrophic consequences. The study proposes a table of principles for evaluation and (re)design of user interface of programs used in highly stressful situations. The principles are grouped based on the human capability hindered by stress. Said principles were evaluated with the help of a case study on a tool for decision support (RASTEP) when facing a nuclear emergency. Statements on usability issues with RASTEP were collected through open group interviews and participant observation and pattern matched with proposed principles. Result of the pattern matching showed the disposition of usability issues per hindered capability following the same pattern as the one of the principles. Furthermore, all but one collected statement were identified as belonging to a group of hindered capabilities and were identified as violating at least one of the proposed principles. This points to the table of principles being able to serve as a tool in evaluating usability of graphic user interface of an existing emergency handling software and identifying its usability issues.
Mycket stressiga situationer kan ha en betydande påverkan på det mänskliga psyket, och det är avgörande att ta hänsyn till när man skapar och utvärderar mjukvara som främst används i pressade situationer. Brist på principer för utvärdering av design av nödhanteringsprogram möjliggör gränssnittslösningar som kan ge upphov till misstag i en miljö där felsteg kan ha katastrofala konsekvenser. Studien föreslår en tabell med principer för utvärdering och (re)design av användargränssnitt för program som används i mycket stressande situationer. Principerna är grupperade efter mänskliga förmågor som hindras av stress. Nämnda principer utvärderades genom en fallstudie av ett verktyg för beslutsstöd (RASTEP) vid kärnkraftsolycka. Vittnesmål om användbarhetsproblem med RASTEP insamlades genom öppna gruppintervjuer och deltagande observation (participant observation), och patternmatchades med de föreslagna principerna. Resultatet av patternmatchningen visade att fördelningen av användbarhetsproblem per hindrad förmåga följde samma mönster som fördelningen av principerna. Dessutom identifierades alla insamlade vittnesmål utom ett tillhöra en grupp av hindrade förmågor och identifierades bryta mot åtminstone en av de föreslagna principerna. Detta tyder på att tabellen med principerna kan användas som ett verktyg för att utvärdera användbarheten av gränssnitt hos ett befintligt nödhanteringsprogram och identifiera dess användbarhetsproblem.

Nuclear fuel performance is highly affected by the behaviour of fission gases, particularly

at elevated burnups, where large amounts of gas are produced and can

potentially be released. The importance of fission gas release was the motivation

for large efforts, both experimentally and theoretically, in order to increase our

understanding of the different steps of the process, and to continuously improve

our models.

Extensions to higher burnups, together with the growing interest in novel types

of fuels such as inert matrix fuels envisaged for the transmutation of minor actinides,

make that one is still looking for a permanently better modelling, based

on a physical understanding and description of all stages of the release mechanism.

Computer simulations are nowadays envisaged in order to provide a better

description and understanding of atomic-scale processes such as diffusion, but even

in order to gain insight on specific processes that are inaccessible by experimental

means, such as the fuel behaviour during thermal spikes.

In the present work simulation techniques based on empirical potentials have

been used, focusing in a first stage on pure uranium dioxide. The behaviour of

point defects was at the core of this part, but also the estimation of elastic and

melting properties.

Then, in a second stage, the study has been extended to the behaviour of helium

and xenon. For helium, the diffusion in different domains of stoichiometry

was considered. The simulations enabled to determine the diffusion coefficient and

the migration mechanism, using both molecular dynamics and static calculation

techniques. Xenon behaviour has been investigated with the additional intention

to model the behaviour of small intragranular bubbles, particularly their interaction

with thermal spikes accompanying the recoil of fission fragments. For that

purpose, a simplified description of these events has been proposed, which opens

perspectives for further work.

/

Les performances du combustible nucléaire sont fortement affectées par le comportement

des gaz de fission, et ce particulièrement lorsqu’un taux d’épuisement

élevé est atteint, puisque d’importantes quantités de gaz sont alors produites

et peuvent potentiellement être relâchées. Les enjeux, entre autre économiques,

liés au relâchement de gaz de fission ont donné lieu à d’importants efforts, tant

sur le plan expérimental que théorique, afin d’accroître notre compréhension des

différentes étapes du processus, et d’améliorer sans cesse les mod`eles. Les extensions

à des taux d’épuisements encore plus élevés ainsi que l’intérêt croissant pour

de nouveaux types de combustible tels que les matrices inertes, envisages en vue

de la transmutation des actinides mineures, font qu’à l’heure actuelle, le besoin

permanent d’une meilleure modélisation, basée sur une compréhension et une description

physique des différentes étapes du processus de relâchement de gaz de

fission, est toujours de mise.

Les simulations par ordinateur ont ainsi été considérée comme un nouvel angle

de recherche sur les processus élémentaires se produisant à l’échelle atomique, à la

fois afin d’obtenir une meilleure compréhension de processus tels que la diffusion

atomique ;mais aussi afin d’avoir accès à certains processus qui ne sont pas observables

par des voies expérimentales, tels que la le comportement du combustible

lors de pointes thermiques.

Dans ce travail, deux techniques, basées sur l’utilisation de potentiels interatomiques

empiriques, ont permis d’étudier le dioxyde d’uranium, dans un premier

temps en l’absence d’impuretés. Cette partie était principalement centrée sur le

comportement des défauts ponctuels, mais a aussi concerné différentes propriétés

élastiques, ainsi que le processus de fusion du composé.

Ensuite l’étude a été étendue aux comportements de l’hélium de du xénon. Pour

ce qui a trait à l’hélium, la diffusion dans différents domaines de stoechiométrie

a été considérée. Les simulations ont permis de déterminer le coefficient de diffusion

ainsi que le mécanisme de migration lui-même. Quant au xénon, outre les

propriétés de diffusion, l’intention fut de se diriger vers la modélisation des petites

bulles intragranulaires, et plus précisément vers leur interaction avec les pointes

thermiques, créées lors du recul des fragments de fission. Une description simplifiée de ce processus a été proposée, qui offre de nouvelles perspectives dans ce

domaine.

Doctorat en Sciences de l'ingénieur
info:eu-repo/semantics/nonPublished

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O progresso de projetos anteriores salientou a necessidade de tratar o problema dos softwares para detecção, a partir de bases de dados de publicações científicas, de tendências emergentes de pesquisa e desenvolvimento. Evidenciou-se a carência de aplicações computacionais eficientes dedicadas a este propósito, que são artigos de grande utilidade para um melhor planejamento de programas de pesquisa e desenvolvimento em instituições. Foi realizada, então, uma revisão dos softwares atualmente disponíveis, para poder-se delinear claramente a oportunidade de desenvolver novas ferramentas. Como resultado, implementou-se um aplicativo chamado Citesnake, projetado especialmente para auxiliar a detecção e o estudo de tendências emergentes a partir da análise de redes de vários tipos, extraídas das bases de dados científicas. Através desta ferramenta computacional robusta e eficaz, foram conduzidas análises de frentes emergentes de pesquisa e desenvolvimento na área de Sistemas Geradores de Energia Nuclear de Geração IV, de forma que se pudesse evidenciar, dentre os tipos de reatores selecionados como os mais promissores pelo GIF - Generation IV International Forum, aqueles que mais se desenvolveram nos últimos dez anos e que se apresentam, atualmente, como os mais capazes de cumprir as promessas realizadas sobre os seus conceitos inovadores.
Dissertação (Mestrado em Tecnologia Nuclear)
IPEN/D
Instituto de Pesquisas Energéticas e Nucleares - IPEN-CNEN/SP

Thermochemical processes for hydrogen production driven by nuclear energy are promising alternatives to existing technologies for large-scale commercial production of hydrogen, without dependence on fossil fuels. In the Copper-Chlorine (Cu-Cl) cycle, water is decomposed in a sequence of intermediate processes with a net input of water and heat, while hydrogen and oxygen gases are generated as the products. The Super Critical Water-cooled Reactor (SCWR) has been identified as a promising source of heat for these processes. In this thesis, the process analysis and simulation models are developed using the Aspen PlusTM chemical process simulation package, based on experimental work conducted at the Argonne National Laboratory (ANL) and Atomic Energy of Canada Limited (AECL). A successful simulation is performed with an Electrolyte Non Random Two Liquid (ElecNRTL) model of Aspen Plus. The efficiency of the cycle based on three and four step process routes is examined in this thesis. The thermal efficiency of the four step thermochemical process is calculated as 45%, while the three step hybrid thermochemical cycle is 42%, based on the lower heating value (LHV) of hydrogen. Sensitivity analyses are performed to study the effects of various operating parameters on the efficiency, yield, and thermodynamic properties. Possible efficiency improvements are discussed. The results will assist the development of a lab-scale cycle which is currently being conducted at the University of Ontario Institute of Technology (UOIT), in collaboration with its partners.
UOIT

This study is intended to investigate and propose a systematic method for uncertainty quantification for the computer code validation application. Uncertainty quantification has gained more and more attentions in recent years. U.S. Nuclear Regulatory Commission (NRC) requires the use of realistic best estimate (BE) computer code to follow the rigorous Code Scaling, Application and Uncertainty (CSAU) methodology. In CSAU, the Phenomena Identification and Ranking Table (PIRT) was developed to identify important code uncertainty contributors. To support and examine the traditional PIRT with quantified judgments, this study proposes a novel approach, the Quantified PIRT (QPIRT), to identify important code models and parameters for uncertainty quantification. Dimensionless analysis to code field equations to generate dimensionless groups (�� groups) using code simulation results serves as the foundation for QPIRT. Uncertainty quantification using DAKOTA code is proposed in this study based on the sampling approach. Nonparametric statistical theory identifies the fixed number of code run to assure the 95 percent probability and 95 percent confidence in the code uncertainty intervals.
Graduation date: 2013
Access restricted to the OSU Community, at author's request, from Dec. 5, 2012 - Dec. 5, 2013

Radioactive xenon gas is a fission product released in the detonation of nuclear devices that can be detected in atmospheric samples far from the detonation site. In order to improve the capabilities of radioxenon detection systems, this work produces [beta-gamma] coincidence spectra of individual isotopes of radioxenon. Previous methods of radioxenon production consisted of the removal of mixed isotope samples of radioxenon gas released from fission of contained fissile materials such as ²³⁵U. In order to produce individual samples of the gas, isotopically enriched stable xenon gas is irradiated with neutrons. The detection of the individual isotopes is also modeled using Monte Carlo simulations to produce spectra. The experiment shows that samples of [superscript 131m]Xe, ¹³³Xe, and ¹³⁵Xe with a purity greater than 99% can be produced, and that a sample of [superscript 133m]Xe can be produced with a relatively low amount of ¹³³Xe background. These spectra are compared to models and used as essential library data for the Spectral Deconvolution Analysis Tool (SDAT) to analyze atmospheric samples of radioxenon for evidence of nuclear events.
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Several computer codes based on one and two-group diffusion theory models were developed for SHUFFLE. The programs were developed to calculate power distributions in a two-dimensional quarter core geometry of a pressurized power reactor. The various coarse-mesh numerical computations for the power calculations yield the following: the Borresen's scheme applied to the modified one-group power calculation came up with an improved power distribution, the modified Borresen's method yielded a more accurate power calculations than the Borresen's scheme, the face dependent discontinuity factor method have a better prediction of the power distribution than the node averaged discontinuity factor method, Both the face dependent discontinuity factor method and the modified Borresen's methods for the two-group model have quite attractive features.
Graduation date: 1992