In accident analysis regarding nuclear power plants, it is very common to use thermal hydraulic system codes, such as TRACE, developed by U.S. NRC. In the case of licensing a power plant, this is one of the necessities. TRACE is a relatively new thermal hydraulic system code and a lot of knowledge is needed to implement it in a correct way, especially in accident analysis, where it is a requirement that the rules and statements in Appendix-K, dealing with criteria for ECCS-models, are modelled. In this thesis an improved model of a Swedish Boiling Water Reactor within TRACE is realized and tested. Afterwards, once a working and representative model has been obtained, a sensitivity study in conducted in order to investigate the sensitivity of TRACE for a couple of thermal hydraulic parameters. The sensitivity study is focussing on the eect of the peak cladding temperature, as well as the coolability of the nuclear fuel in terms of quenching and quench-front velocities. It is found to be hard to say unilaterally what the eect of changing a certain number of parameters on the reactor behaviour is. As it turns out to be, although strongly related, the peak cladding temperatures and the quench phenomena can behave dierently

Les travaux menés au cours de cette thèse s’inscrivent dans le cadre général de l’évaluation du terme source (relâchement de gaz de fission) en situation incidentelle de type APRP (Accident de Perte de Réfrigérant Primaire). Lors de tels transitoires thermiques, le relâchement de gaz de fission se fait par bouffées successives : une première entre 600°C et 800°C et la seconde à environ 1100°C. Ces travaux de thèse s’intéressent à cette première. Il semblerait que la bouffée à 600-800°C proviendrait du centre de la pastille combustible. L’objectif de cette thèse est d’étudier les mécanismes à l’origine de cette bouffée.Afin de mieux comprendre ces mécanismes, une étude a été menée sur l’ensemble des traitements thermiques réalisés dans la plateforme expérimentale MERARG. L’analyse de cette base de données a révélé 2 points importants : 1) Dans les conditions expérimentales de MERARG, aucune fagmentation significative du combustible n’est observée à des températures inférieures à 1000°C. 2) Le niveau de relâchement de gaz de fission entre 600°C et 800°C semble augmenter avec le temps d’entreposage du combustible.Le premier point indique que la fragmentation du combustible n’est pas une condition nécessaire au relâchement de gaz de fission dans cette gamme de température : d’autres mécanismes peuvent être à l’origine de ce relâchement. Durant l’entreposage, le combustible est soumis principalement à l’auto-irradiation α. Celle-ci a pour effet de créer des défauts dans une zone qui n’en contenait initialement pas. Nous avons démontré que la cinétique du relâchement de gaz de fission entre 600°C et 800°C est concomitante avec la cinétique de recuit de défauts d’autoirradiation α. De plus, une cinétique auto-catalytique de germination-croissance de nano-clusters de gaz a été développée et confrontée aux résultats expérimentaux. En outre, une étude sur matériaux simulants démontre que, sur des pastilles d’UO2 frittées et implantées en xénon, une irradiation en régime électronique a pour effet d’accroitre le relâchement entre 600°C et 800°C. La littérature décrit la remise en solution des bulles de gaz de fission sous l’effet d’une irradiation de ce type. De plus, lors de leur remise en solution, les gaz de fission s’insèrent dans les défauts de la structure cristalline. Lors d’un traitement thermique, le recuit des défauts entraine la mobilité des atomes de gaz de fission insérés de ces mêmes défauts. Par germination-croissance, les paires gaz/défauts rejoignent un chemin de sortie, les gaz de fission sont donc relâchés.Ce travail a donc permis de retenir l’hypothèse d’un mécanisme de relâchement de gaz de fission entre 600°C et 800°C par recuit de défauts sans fragmentation significative du combustible. / The subject of this thesis concerns the evaluation of the source term (fission gas release) in incidental situations of type LOCA (Loss of Coolant Accident) of nuclear fuels. During such thermal transients, the fission gas release is characterized by successive bursts : the first one occurring between 600 and 800°C, and a second one at about 1100°C. This work is about the first burst release. It appear that this one come from the centre of the fuel pellet. The aim of this thesis is to study the mechanisms responsible for the fission gas release between 600°C and 800°C.To this purpose, we collected more than 200 results of thermal treatments carried out using the experimental platform MERARG. The analysis of this database reveals two important results : under the experimental conditions of MERARG, no significant fragmentation of the fuel was observed at temperatures below 1000°C ; the amount of fission gas release between 600°C and 800°C appears to increase with fuel storage time.The first result suggests the fragmentation of the fuel is not a necessary condition for the release of fission gas in this temperature range. Other mechanisms may then be responsible for this gas release. During its storage, the fuel undergoes α particle self-irradiation. We demonstrate that the kinetics of fission gas release between 600°C and 800°C is simultaneous with the kinetics of the annealing of self-irradiation defects at this same temperature. The mechanism involves an autocatalytic process leading to a kinetic of fast germination-growth of gas nano-clusters. This model perfectly explains the experimental results in the database. To confirm this mechanism, a study on surrogate materials demonstrates that, in UO2 pellets sintered and implanted by Xe, irradiations in the electronic regime actually promote the release of implanted Xe at those temperatures. The re-dissolution of the fission gas bubbles by this kind of irradiation is consistent with observations in other contexts. Those conclusions allow to extend the mechanism for release to irradiated fuel.During the storage of the fuel, α self-irradiation promotes the re-dissolution of the trapped gas. The consequences of this effect are particularly important in the region close to the grain boundaries, where the concentration of defects is also larger. The irradiation mechanism increases the fraction of fission gas available for release, depleting the amount of gas initially trapped in bubbles. The gas in solution can effectively be carried by crystal defects, largely available in the irradiated fuel and whose migration at 600-800°C induces the mobility of the fission gas. When they reach an outlet path, the gas can be released from the pellet in a way consistent with the model of autocatalytic germination-growth we developed to explain the macroscopic results of the database.In conclusion, this work supports the hypothesis of a mechanism of fission gas release in the range 600-800°C via a mechanism involving the migration and annealing of defects without significant fragmentation of the fuel.

As a latest developed computational code, TRACE is expected to be useful and effective for analyzing the thermal-hydraulic behaviors in design, licensing and safety analysis of nuclear power plant. However, its validity and correctness have to be verified and qualified before its application into industry. Loss-of-coolant accident (LOCA) is a kind of transient thermal hydraulic event which has been emphasized a lot as a most important threat to the safety of the nuclear power plant. In the present study, based on FIX- II LOCA tests, simulation models for the tests of No. 3025, No. 3061 and No. 5052 were developed to validate the TRACE code (version 5.0 patch 2). The simulated transient thermal-hydraulic behaviors during the LOCA tests including the pressure in the primary system, the mass flow rate in certain key parts, and the temperature in the core are compared with experimental data. The simulation results show that TRACE model can well reproduce the transient thermal-hydraulic behaviors under different LOCA situations. In addition, sensitivity analysis are also performed to investigate the influence of particular models and parameters, including counter current flow limitation (CCFL) model, choked flow model , insulator in the steam dome, K-factor in the test section, and pump trip, on the results. The sensitivity analyses show that both the models and parameters have significant influence on the outcome of the model.

<p> In a postulated critical break LOCA in a CANDU reactor it is possible that heatup of a pressure tube (PT) causes ballooning contact with the calandria tube (CT). Stored heat in the PT is transferred out, yielding a high PT-CT heat flux, which can cause dry out of the CT and establishment of pool film boiling on the outer surface of the tube. The safety concern associated with this condition is that if the temperature of the CT experiencing film boiling gets sufficiently high then failure of the fuel channel may occur. However, quench heat transfer can limit the extent and duration of film boiling as has been experimentally observed. Current estimates of quench temperatures during pool film boiling are based primarily on experimental correlations. In this dissertation a novel mechanistic model of pool film boiling on the outside of a horizontal tube with diameter relevant to CT (approximately 130 mm) has been developed. The model is based in part upon characterizing the vapor film thickness for steady state film boiling under buoyancy driven natural convection flows around a tube located horizontally in a large liquid pool. Variations in steady state vapor film thickness as a function of the incident heat flux, the temperature of the CT outer wall, and the subcooling of the bulk liquid are analyzed. The calculated effective film boiling heat transfer coefficient is compared to available experimental data. Finally a transient equation is developed which quantifies the instability of the vapor film and a possible occurrence of rapid quench when a step change in governing parameters occurs, such as liquid subcooling. This mechanistic model can be employed in safety analysis to demarcate the conditions under which fuel channel failure will not occur in a postulated critical break LOCA.</p> / Thesis / Master of Applied Science (MASc)

When soluble boron is used to control reactivity, there are two particular events which can challenge long-term core cooling (LTCC) during the small break loss-of-coolant accident (SBLOCA): boron precipitation and boron dilution. The initial consequences of the SBLOCA are mitigated by the emergency safety systems, but the core continues to boil. As boron is less volatile than steam, the steam is virtually boron free. All the boron remains in the core, the boron concentration in the core rises. If the solubility limit is reached, precipitation could occur. The boron precipitation event was historically considered to be bounded by the large break accident. However, there are characteristics of the SBLOCA which cannot be neglected and an SBLOCA specific methodology is required. On the opposite end of the boron concentration spectrum is the SBLOCA boron dilution event. The steam generators remove heat from the primary and condense the steam. The condensation of the boron-free steam can result in the accumulation of a deborated slug of water. If natural circulation restarts, the slug can be transported toward the core and potentially reduce the core boron concentration enough to induce a recriticality. This thesis describes two analytical methodologies for these SBLOCA LTCC events. The two methodologies have a similar approach. Both use transient system analyses for inputs to and justification of the follow-on boron concentration calculations. For boron precipitation, a maximized concentration is calculated with the Small Break Boron Precipitation model. For boron dilution, a minimized core inlet concentration is calculated using computational fluid dynamics. / Master of Science

This dissertation covers the application of heat accumulation systems in nuclear power engineering, namely in nuclear power plants. It is mainly a case of passive emergency systems, whose task is to accumulate the heat produced in the reactor’s active zone and in spent fuel pools during DBA (design-basis accidents) or beyond DBA. A particular example of heat accumulation is steam condensation after LOCA (loss of coolant accident). The primary circuit steam leakage increases containment pressure and has to be decreased by the steam condensation. This thesis deals with a theoretical substitute for ice condensers, which are used as a passive safety measure in some nuclear power plants. The substitute involves a choice of an alternative material, whose melting temperature (for heat accumulation) is closer to nuclear power plant operating temperatures. The other part of the dissertation discusses heat accumulation in spent fuel pools in case of all cooling systems failure.

Submitted by (lucia.rodrigues@ufrpe.br) on 2016-12-12T15:20:24Z No. of bitstreams: 1 Rosa Maria Saraiva.pdf: 1981088 bytes, checksum: 280097755285a21a00ad5a0f584b6d4c (MD5) / Made available in DSpace on 2016-12-12T15:20:24Z (GMT). No. of bitstreams: 1 Rosa Maria Saraiva.pdf: 1981088 bytes, checksum: 280097755285a21a00ad5a0f584b6d4c (MD5) Previous issue date: 2008-12-12 / This study aimed to analyse the strategies of communication used by the Program of Combat to the Rural Poverty, of the Renascer Project, and for the Peixe de Rede Project, of the Sustainable Development Program in Pernambuco Forest Zone (PROFOREST), in the communities Atapuz, Barra de Catuama and Tejucupapo, Goiana district, Pernambuco. It this case study of qualitative nature, supported of the methodologist proceedings used by the institutions, quoted above, through institutional documents reading and interview semi - structuralized, distributed by categories for analysis. The study had as theoretical concepts scientific boarding in the aspect of the Fishing Extension, local Development and Communication. We base this study in the research problem: which results gotten by service of Fishing Extension of the Renascer and Promata Projects in the communities? And which relation of these results with the thematic of the communication strategies for the local development? The research counted on a amostral universe of twenty local social actors, five technician and two managers of the Renascer and Promata programs. There was used individual interviews carried out with carving and photographic register, to the field work The analysis of the content, applied to the data obtained through the interviews made possible to visualize like the strategies of communication they were used for the development of these projects. The communication strategies appear very incipient, it still favoring punctual results what does not contemplate the real necessities of these communities that use fishing activities guarantee of better conditions of life for the community and future generations. / Esta pesquisa objetivou analisar as estratégias de comunicação utilizadas pelo Programa de Combate à Pobreza Rural, do Projeto Renascer, e pelo Projeto Peixe de Rede, do Programa de Apoio ao Desenvolvimento Sustentável da Zona da Mata de Pernambuco (Promata), nas comunidades de Atapuz, Barra de Catuama e Tejucupapo, Município de Goiana, Pernambuco. Trata-se de um estudo de caso, de natureza qualitativa, apoiado em procedimentos metodológicos utilizados pelas instituições acima citadas, por meio de consulta a documentos institucionais e entrevistas semi- estruturadas, distribuídas por categorias para análise. Para tanto, o estudo contou com o aporte teórico no que diz respeito à Extensão Pesqueira, Desenvolvimento local e Comunicação. Embasamos este estudo na seguinte questão: quais os resultados obtidos por meio de uma terceirização de serviço de Extensão Pesqueira do Projeto Renascer e Promata nas comunidades e qual a relação desses resultados com a temática da estratégias de comunicação para a desenvolvimento local? A pesquisa contou com um universo amostral de vinte atores sociais locais, cinco técnicos e dois gestores dos programas Renascer e Promata. Para a pesquisa de campo, foram utilizadas entrevistas individuais realizadas com gravação em áudio. A análise dos dados possibilitou visualizar como as estratégias de comunicação foram utilizadas para o desenvolvimento desses projetos. As estratégias de comunicação aparecem de maneira muito incipiente, favorecendo resultados pontuais que ainda não contemplam às reais necessidades dessas comunidades que se utilizam das atividades pesqueiras para garantir melhores condições de vida para a comunidade e as futuras gerações.

Thesis (Ph. D. (Public Administration)) -- University of Limpopo, 2016 / Municipalities have the responsibility to deliver services to the communities in a fast and efficient manner, and to deliver these services there is a need for financial resources and institutional capacity. The challenge that beset the municipalities is that they struggle to collect revenue. There is a gap between available financial resources and the municipal expenditure needs largely as a result of the revenue collection challenges facing the municipalities. The revenue collection challenge therefore, needs to be adequately addressed for the municipalities to be successful. Municipalities have the right to finance their affairs through charging fees for services; imposing surcharges on fees, rates, levies and duties. The municipal council have the responsibility to implement and adopt tariff policies. These tariff policies must espouse the principles that ensure the equitable treatment of municipal service users. Tariff policies must also ensure that the amount paid by individual users for services is proportionate to their usage. Municipalities should, in terms of law, differentiate between the different categories of ratepayers, users of services, debtors, taxes, services and service standard. This study was based on the combination of both qualitative and quantitative research design. It followed a case study approach of comparative investigation between the Tshwane Metropolitan Municipality revenue collection and South African Revenue Service. The quantitative research was done through the administration of questionnaires to the ratepayers, corporate taxpayers and tax practitioners. This was complemented by the qualitative in-depth interview questionnaires administered to both South African Revenue Service and the Tshwane Metropolitan municipality to ensure that the data collected are both collaborated and diverse to enable the researcher to draw a balanced conclusion. In the light of the above, this study therefore investigated the strategies and legislative framework that is employed by the municipalities and contrasted with those that are used by the South African Revenue Service, with the aim of taking possible learnings that can be applied in the municipalities. The study concluded by proposing guidelines that can be used by municipalities in revenue collection.

The Thesis discusses about safety systems of nuclear power plants, primarily about passive containment protection. That have aims to reduce increase of pressure during the LOCA. It occupy specific type of passive system, it is Ice Condenser. In chapters is make step by step design of replacement current material for other material, that changes its phase at higher temperature. It chose a suitable material, prepared thermal calculation and structural design. Finally is prepared the economic analysis of production of new type condenser.

This thesis studied the performance of the thermal hydraulic software TRACE applied on an intermediate sized break (IB) happening on the cold leg in a pressurized water reactor (PWR), causing a loss-of-coolant accident (LOCA). The same accident has previously been simulated in the PKL Test Facility, which is a scaled version of a PWR and is used to simulate transients stemming from different accidents. The thesis was performed as a semi-blind study: firstly, the accident was simulated without any knowledge of the PKL results. When a final blind model was chosen, the PKL results were revealed, and the TRACE model was improved. Before the simulations of the IB-LOCA took place, the new internal parts in the upper parts of the reactor pressure vessel in PKL had to be modelled, and the steady state had to be tuned to attain the correct initial conditions. The simulations were performed by using the software SNAP together with TRACE, providing a graphical interface. TRACE achieved steady state with satisfying results regarding water levels, pressure losses and mass flows. The temperatures in TRACE deviated from the PKL temperatures but an explanation is uncertainties in PKL. To verify TRACE’s core output power, the calculation of the power was done by using mass flow rate and specific entropy and comparing to the heaters’ specified power. This resulted in lower output power meaning that the coolant was not heated enough. This indicated non-physical energy losses in the TRACE model and should be further investigated.The blind transient simulation, modelled with default choked flow and no offtake model, resulted in TRACE overestimating the break mass flow and the peak cladding temperatures, compared to the PKL reference solution. This resulted in the pressure decreasing too quickly and too early activation of the safety system. The modified simulations showed that it is important that the offtake model, which accounts for different flow regimes, is activated. Default choked flow multipliers were the multipliers that performed the best. However, none of the transient simulations could be completed due to fatal errors and memory problems, but some conclusions could be drawn from the observed trends. This concluded in the offtake model being most important due to stratified flow occurring.