The Overhead Line Equipment (OLE) is a critical sub-system of the 25kV AC overhead railway electrification system, which is the main method of railway electrification on the British railway network. OLE failures can result in significant delays and pose risks to passenger safety, therefore, inspection and maintenance is undertaken to improve component reliability and uphold the availability of the system. OLE asset management strategies can be evaluated using a life cycle cost analysis that considers degradation processes and maintenance activities of the OLE components. The investment required to deliver the level of performance desired by railway customers and regulators can be based on evidence from the analysis’ results. This thesis presents a methodology for modelling the asset management and calculating the whole life cost of the OLE to allow such analysis to take place. This research has developed a High Level Petri net model to simulate the degradation, failure, inspection and maintenance of the main OLE components in a stochastic manner. The model simulates all the main OLE components concurrently in the same model and fixed time interval inspections and condition-based maintenance regimes are considered. The various dependencies between the different components and processes considered, such as opportunistic inspection and maintenance, are also taken into account. The use of High Level Petri nets allows the processes considered to be modelled in a more accurate and efficient manner in comparison to standard Petri nets. The model is used to calculate various statistics associated with the cost, maintenance requirements and reliability of the individual OLE components and the OLE system over its life cycle. This is demonstrated using an example analysis for a 2-mile section of electrified line, which also describes how the outputs obtained can be used by decision makers to study the performance of the components and the implications of the maintenance strategy evaluated by the model. Finally, a Genetic Algorithm is used in conjunction with the Petri net developed to find the optimum maintenance strategies that result in the lowest total cost of the system. The optimum strategy chosen results in a 15% lower expected total cost and 10% fewer expected failures in comparison to the maintenance strategy currently implemented for the OLE on the British railway network, whilst requiring a similar number of maintenance visits. The methodology presented considers the OLE components and the processes described above in more detail than previous literature associated with asset management and life cycle cost analysis of the OLE. Additionally, the suitability and ways in which Petri nets can be used for modelling the asset management of other large engineering systems, comprised of numerous components with various dependencies, is confirmed. Furthermore, the practical use of the model, as an asset management tool, capable of calculating a comprehensive range of outputs calculated, is demonstrated.

No description available.

The safe transportation of new and spent nuclear fuel is an essential part of the nuclear fuel cycle. The aim of this thesis was to obtain a more thorough understanding of the mechanical loading acting on heavy nuclear packages during rail transportation. There were two motivating factors for this study. Firstly, the design of equipment used to tie down a package to its conveyance has become more challenging with the recent trend of increasing package mass; often exceeding 100 tonnes. This difficulty is due to the advisory acceleration factors recommended for design. Despite widespread acceptance that the factors ensure safety, it is also recognised that for heavier packages they can be prohibitive and result in over engineered tie down systems. Secondly, transportation imparts complex dynamic mechanical loading on packages and the fuel assemblies within them. There have been no reported instances in the UK of problems caused by fuel vibrations. However international studies have prompted this investigation. A rail wagon and tie down system for a 100 tonne package were instrumented with accelerometers and strain gauges. The measurements were taken during a routine rail journey from Barrow-in-Furness to Sellafield. Continuous data was digitally recorded with a sampling rate sufficient to capture shock and vibrations up to 100 Hz. Accelerometers were selected to measure very low frequencies to capture quasi-static loading. Investigation of the frequency content of the accelerations indicated that digital filtering of the data is necessary to determine the magnitudes of the structural loading on tie downs. A method for designing a suitable filter has been developed. A sensitivity analysis of different filters indicated there is a possibility for over estimating loads based on measured data due to poor filter design. Industrial design of tie downs using FEA requires pragmatic run times. This motivated a comparison of the measured strain time histories with the results of a linear static FEA model. The correlation between measured and predicted strains, was strong at frequencies < 3.5 Hz. A residuals analysis indicated that the model predicted the underlying strain process accurately. The methods described are generic and adaptable. They will aid any future experimental work, to characterise shock, vibration and quasi-static loads acting on nuclear packages and their ancillary equipment.

This master thesis focuses on the possible connection between the digital television broadcasting DVB and the Twitter social network. The target platform is the Hybrid Broadcast Broadband TV platform, which combines television broadcasting with a data received from broadband. The created system is composed of a HbbTV application and a server, that connects the application with the Twitter and searchs for additional data about the current video content. Usage of the resulting solution in real television broadcasting could potentially increase the amound of tweets related to the television broadcasting, increase the knowledge about HbbTV technology and attrack young generation of viewers. All that can result into slight increased number of viewers.

This master’s thesis deals with using Text mining as a method to predict tags of articles. It describes the iterative way of handling big data files, parsing the data, cleaning the data and scoring of terms in article using TF-IDF. It describes in detail the flow of program written in programming language Python 3.4.3. The result of processing more than 1 million articles from Wikipedia database is a dictionary of English terms. By using this dictionary one is capable of determining the most important terms from article in corpus of articles. Relevancy of consequent tags proves the method used in this case.

This thesis deals with the problem of mining structured information concerning the features of the products from the open text, using open information extraction. These features will make it easier for customers to choose their product. In the beginning, it deals with existing solutions, their shortcomings and analysis of available systems for open information extraction. Furthermore, the theoretical background and technology used in the creation of the system, the design of the system itself and its implementation are discussed. At the end, the system testing, its results and extensions that could be implemented in the future are described.

Detecting topics by extracting keywords from written text using TF-IDF has been studied and successfully used in many applications. Adding a semantic layer to TF-IDF-based topic detection using WordNet synonyms and hypernyms has been explored in document clustering by assigning concepts that describe texts or by adding all synonyms and hypernyms that occurring words have to a list of keywords. A new method where TF-IDF scores are calculated and WordNet synset members’ TF-IDFscores are added together to all occurring synonyms and/or hypernyms is explored in this paper. Here, such an approach is evaluated by comparing extracted keywords using TF-IDF and the new proposed method, SynPlusTF-IDF, against manually assigned keywords in a database of scientific abstracts. As topic detection is widely used in many contexts and applications, improving current methods is of great value as the methods can become more accurate at extracting correct and relevant keywords from written text. An experiment was conducted comparing the two methods and their accuracy measured using precision and recall and by calculating F1-scores.The F1-scores ranged from 0.11131 to 0.14264 for different variables and the results show that SynPlusTF-IDF is not better at topic detection compared to TF-IDF and both methods performed poorly at topic detection with the chosen dataset.

The current study focuses on improving and testing the CTF thermalhydraulics computer code. CTF is a thermalhydraulic code used for subchannel analysis of nuclear power reactors developed as part of the US DOE CASL program and distributed by North Carolina State University. Subchannel analyses are used to predict the local fuel temperatures and coolant conditions inside a complex nuclear fuel assembly. Such calculations are used to improve designs of nuclear fuel, improve operating margins, or perform safety analysis. An important part of the code development process is the verification and validation for its intended use. In this work validation activities are performed using the RISO experiments are modeled in CTF for adiabatic and diabatic cases in annular flow regimes and a limited set of tests in CANDU geometries. The CTF predictions significantly overpredicted the pressure drop for cases involving annular flow conditions. Depending on the application, such overprediction can result in significant errors in the computation of fuel element dryout and other figures of merit. For example, an analysis using fixed pressure boundary conditions CTF predicts much lower subchannel flows and hence fuel element temperatures may be overestimated. On the other hand, for a scenario with mass flux and inlet pressure as boundary conditions, the impact of pressure drop discrepancies on dryout predictions may be lower. Therefore, there is a particular focus in this thesis on the two-phase pressure drop models and the RISO experiment specifically, since the RISO tests involve a range of annular flow conditions which is prototypical of many CANDU accident analysis conditions. In addition to the RISO experiments, 28-element CANDU full scale rod bundle experiments are modeled in CTF for single-phase and two-phase flow conditions. Cases are modeled for crept and uncrept conditions with different bearing pad heights i.e., 1.17 mm and 1.35mm. Pressure drop predictions are compared with the experimental results where single-phase comparisons are in good agreement while an overprediction of ~25% is observed for two-phase conditions. The effect of bearing pads on the subchannel local parameters, like mass flow rate, are also studied. Furthermore, the effect of turbulent mixing rate on subchannel enthalpy distribution in the bundle and CHF in different subchannels is also analyzed. Based on the comparison to the RISO and CANDU 28 element test databases, the overprediction of pressure drop in the annular flow regime needs improvement in the current version of CTF. This overprediction of the frictional pressure drop results from either wall drag or interfacial shear stress phenomena. In this study, it is demonstrated that the issue occurs mostly as a result of interfacial friction factor modelling this work examines several alternative approaches. The results show the Ju’s and Sun’s interfacial friction factor better predicts the results among all the other six correlations implemented in CTF. The major impediment in further testing of CTF is that it lacks the capability to simulate R-134a fluids. Given there is a large database of R-134a two-phase tests, another aspect of this thesis is to extend CTF for application and validation using refrigerants. The current CTF version only supports fluid properties for water and FLiBe salts. By adding R-134a fluid properties the testing and validation range of CTF is broadened for different experiments performed using R-134a fluids. CHF experiments are modeled in CTF and results are compared with experimental data. For local conditions correlation, 2006 water LUT are used to predict CHF and DNBR. The fluid-to-fluid scaling method is applied in CTF when using CTF with R-134a fluid properties for CHF and DNBR predictions to account for the difference in fluid properties between R-134a and the CHF look-up table. / Thesis / Master of Applied Science (MASc) / COBRA-TF (CTF) is a thermalhydraulic code, based on the historical code COBRA-TF, used for subchannel analysis of nuclear power reactors. Subchannel analysis can be used to predict the local fuel temperatures and coolant conditions inside a complex nuclear fuel assembly. CTF is a transient code that simultaneously solves conservation equations for mass, momentum, and energy for the three coolant phases present, i.e. vapor, continuous liquid, and entrained liquid droplet phases. The scope of the current study includes 1) testing the code for conditions relevant to CANDU accident analysis, 2) refinement of the models that are used in two-phase interfacial friction calculations, and 3) inclusion of alternate fluid properties. The testing of CTF is performed with different experimental databases covering CANDU thermalhydraulic conditions. The refinement is done by improving the pressure drop prediction in the annular flow regime by using different interfacial friction factor correlations from earlier studies in the literature. The current CTF version includes water and liquid salt properties (FLiBe) for coolant fluids. Freon (R-134a) fluid properties have been added in CTF in order to broaden the testing range of CTF for different experimental database using R-134a as working fluid.

Den minimalinvasiven, kathetergestützten Verfahren der Aortenklappenimplantation (TA-AVI und TF-AVI) kommt aufgrund der Zunahme der Hochrisikopatienten im demographischen Wandel in der Behandlung der hochgradigen Aortenstenose eine immer größere Bedeutung zu. Auch wenn der konventionelle Aortenklappenersatz immer noch als Goldstandard in der Therapie der Aortensklerose gilt, erweist sich die katheterinterventionelle Methode insbesondere für Patienten mit diversen Komorbiditäten oder der sogenannten Porzellanaorta als eine sehr gute und allseits anerkannte Alternative. Wir wollten in dieser Studie erstmalig die Embolielast untersuchen, die während einer kathetergestützten transapikalen (TA-AVI) oder transfemoralen (TF-AVI) Aortenklappenimplantation auftritt und diese mit dem konventionellen Aortenklappenersatz (AVR) vergleichen. Unser Patientenkollektiv bestand dabei aus Hochrisikopatienten, wobei die TA-AVI Gruppe das signifikant höhere Risiko aufwies. Unsere Ergebnisse zeigten einerseits eine Abhängigkeit der Embolielast von diversen Arbeitsschritten während der verschiedenen Formen der Aortenklappenimplantation (TA-AVI, TF-AVI und AVR). Zudem konnten wir ein geringeres Embolierisiko für die Gruppe der TA-AVI verglichen mit der TF-AVI und AVR aufzeigen. In den MRT-Untersuchungen zeigten sich postinterventionell ischämische cerebrale Läsionen, die jedoch nicht mit den Ergebnissen der neurokognitiven Testverfahren korrelierten.:1. Einleitung 1 1.1. Aortenklappenstenose 1 1.1.1. Definition 1 1.1.2. Ätiologie und Epidemiologie 2 1.2. Therapeutische Verfahren 3 1.2.1. Konventioneller Aortenklappenersatz 5 1.2.2. Entwicklung der kathetergestützten Aortenklappenimplantationsverfahren 5 1.3. Transkranielle Dopplersonographie (TCD) 7 1.3.1. Grundlagen des Ultraschalls 7 1.3.2. Doppler-Effekt 8 1.3.2.1. Continuous-Wave-Doppler (CW-Doppler) 9 1.3.2.2. Pulsed-Wave-Doppler (PW-Doppler) 9 1.3.3. High Intensitiy Transient Signals (HITS) 9 1.4. Zielstellung der Arbeit 10 2. Material und Methoden 12 2.1. Studienort und -zeitraum 12 2.2. Patientenkollektiv und Einschlusskriterien 12 2.3. Therapeutische Verfahren 13 2.3.1. Konventioneller Aortenklappenersatz 13 2.3.2. Kathetergestützte Aortenklappenimplantation 13 2.3.2.1. Klappentypen und Implantationssysteme 15 2.3.2.2. Transfemorale Implantation der CoreValve® Prothese 17 2.3.2.3. Transapikale Implantation der Edwards SAPIEN® Prothese 18 2.4. Prä- und postinterventionelle Untersuchungen 21 2.4.1. MRT 21 2.4.2. Protein S100𝛽-Bestimmung 22 2.4.3. Neurokognitive Testverfahren 22 2.4.3.1. Verbaler Lern- und Merkfähigkeitstest (VLMT) 23 2.4.3.2. Trail Making-Test A und B 24 2.4.3.3. Grooved Pegboard 25 2.4.3.4. Canadian Neurological Scale 26 2.5. Periinterventionelle transcranielle Dopplersonographie 26 2.5.1. Untersuchungsverfahren 26 2.5.2. Gerätebeschreibung 28 2.5.3. Unterteilung der HITS 30 2.6. Statistik 30 3. Ergebnisse 32 3.1. MRT 32 3.2. Protein S100𝛽 34 3.3. Neurokognitive Testverfahren 36 3.3.1. Verbaler Lern- und Merkfähigkeitstest 36 3.3.2. Trail Making-Test A und B 41 3.3.3. Grooved Pegboard 43 3.3.4. Canadian Neurological Scale 46 3.4. Periinterventionelle transkranielle Dopplersonographie 48 4. Diskussion 59 4.1. Hintergrund 59 4.2. Bewertung der postinterventionellen craniallen MRT-Befunde 60 4.3. Analyse postinterventioneller Protein S100𝛽 -Spiegel im Serum 61 4.4. Beurteilung der neurokognitiven Testverfahren 62 4.5. Bewertung der transkraniellen Dopplersonographie-Befunde der TA-AVI, TF-AVI und AVR 63 4.6. Schlussfolgerung und Ausblick 65 5. Zusammenfassung der Arbeit 69 6. Literaturverzeichnis 72 7. Anlagen 82 8. Erklärung über die eigenständige Abfassung der Arbeit 95 9. Curriculum Vitae 96 10. Danksagung 98

In recent years, in parallel with advances in computer technology, a multitude of computer tools have been developed through which it is possible to obtain a detailed description of the phenomena occurring in the core of nuclear reactors. The final ob-jective of these new tools is to perform safety analysis using best estimate techniques. The best estimate techniques, as opposed to the conservative ones, allow the operation of the reactor with narrower safety margins, and thus greater core economy. In this context, in this work is developed an multiphysics computer platform that inte-grates simulation codes that cover most of the physics that take place in nuclear reac-tors. For the integration of the different feedback phenomena between thermal-hydraulics, neutronics and heat transfer, a series of couplings have been developed between the codes that compose the platform. All the developments carried out are intended to realistically represent the design and behavior of the nuclear facility, in-cluding the control system, fuel elements and fuel rods. The computer platform includes some of the state-of-the-art codes for reactor analysis. The thermal-hydraulics is covered with a developed coupled code, consisting of the semi-implicit coupling between the TRACE system code and the subchannel code COBRA-TF (CTF), whose parallel version has been created in this work. In transients where three-dimensional neutron calculations are necessary, the explicit coupling be-tween the three-dimensional PARCS core simulator and the subchannel code CTF has been developed. For the analysis of the integrity of the fuel rods, the FRAPCON and FRAPTRAN codes are used, coupling the latter explicitly with CTF. All the developed tools have been included in the same computer platform that en-compasses and coordinates the simulations under the user's guidelines. The platform has enough flexibility to perform safety studies in a multitude of operational or acci-dental scenarios, and it is hoped that in the future it may be used for supporting li-cense calculations. The developed tools have been verified through a series of practical applications in different transient and accidental scenarios in light water reactors. The results obtained have been compared with actual plant measurements and with the results of other simulation codes showing adequate predictive capacity. The work carried out in this doctoral thesis is part of the research line financed by the Ministerio de Economía y Competitividad in the NUC-MULTPHYS project (ENE2012-34585) and the interdisciplinary collaboration projects of the Universitat Politècnica de Valencia COBRA_PAR (PAID-05-11-2810) and Open-NUC (PAID-05-12). / En los últimos años, paralelamente a los avances en tecnología informática, se están desarrollando multitud de herramientas informáticas mediante las que es posible obte-ner una descripción detallada de los fenómenos que tienen lugar en el núcleo de los reactores nucleares. El objeto de estas nuevas herramientas es el de realizar análisis de seguridad en reactores nucleares utilizando técnicas de mejor estimación. Las técnicas de mejor estimación, en contraposición con las conservadoras, permiten la operación del reactor con márgenes de seguridad más estrechos, y por tanto mayor economía del núcleo. En este contexto, en la presente tesis doctoral se desarrolla una plataforma informática que integra códigos informáticos que cubren la mayor parte de las físicas que tienen lugar en los reactores nucleares. Para la integración de los diferentes fenómenos de realimentación entre termohidráulica, neutrónica, mecánica y transmisión de calor se han desarrollado una serie de acoplamientos entre los códigos que componen la plata-forma. Todos los desarrollos realizados tienen por objetivo representar de forma rea-lista el diseño y comportamiento de la instalación nuclear, incluyendo el sistema de control, los elementos y las varillas de combustible. En la plataforma informática se incluyen algunos de los códigos de última generación (estado de arte) para el análisis del comportamiento de reactor. En el plano termohi-dráulico se utiliza el código acoplado desarrollado, formado por el acople semi-implícito entre el código de sistema TRACE y el de subcanal COBRA-TF (CTF), cuya versión paralela ha sido creada en este trabajo. En transitorios en los que resultan ne-cesarios los cálculos de neutrónica tridimensional, se ha desarrollado el acople explíci-to entre el simulador tridimensional de núcleos PARCS y el código de subcanal CTF. Para el análisis de la integridad de las varillas de combustible se emplean los códigos FRAPCON y FRAPTRAN, acoplando este último de forma temporalmente explícita con CTF. Todos los desarrollos realizados se han incluido en una misma plataforma informática que los engloba y coordina las simulaciones bajo las directrices del usuario. La plata-forma posee suficiente flexibilidad para realizar estudios de seguridad en multitud de escenarios operacionales o accidentales, y se desea que en un futuro pueda ser utilizada en cálculos de apoyo a licencia. Las herramientas desarrolladas han sido verificadas mediante una serie de aplicaciones prácticas en distintos transitorios y escenarios acci-dentales en reactores de agua ligera. Los resultados obtenidos se han comparado con medidas reales de planta y con los resultados de otros códigos de simulación mostran-do una adecuada capacidad predictiva. El trabajo realizado en la presente tesis doctoral se enmarca dentro de la línea de in-vestigación financiada por el Ministerio de Economía y Competitividad en el proyec-to NUC-MULTPHYS (ENE2012-34585) y los proyectos de colaboración interdisci-plinar de la Universitat Politècnica de Valencia COBRA_PAR (PAID-05-11-2810) y Open-NUC (PAID-05-12) / En els últims anys, paral·lelament als avanços en tecnologia informàtica, s'estan desenvolupant multitud de ferramentes informàtiques mitjançant les quals és possible obtindre una descripció detallada dels fenòmens que tenen lloc en el nucli dels reactors nuclears. L'objecte final d'aquestes noves ferramentes és el de realitzar anàlisis de segu-retat a reactors nuclears utilitzant tècniques de millor estimació. Les tècniques de mi-llor estimació, en contraposició amb les conservadores, permeten l'operació del reactor amb marges de seguretat més estrets, i per tant una major economia del nucli. En aquest context, en el present treball de tesi es desenvolupa una plataforma in-formàtica que integra codis informàtics que cobreixen la major part de les físiques que tenen lloc als reactors nuclears. Per a la integració dels diferents fenòmens de reali-mentació entre termohidràulica, neutrònica i transmissió de calor s'han desenvolupat una sèrie d'acoblaments entre els codis que componen la plataforma. Tots els desenvo-lupaments realitzats tenen per objectiu representar de forma realista el disseny i com-portament de la instal·lació nuclear, incloent el sistema de control, els elements i les varetes de combustible. A la plataforma informàtica s'inclouen alguns dels codis d'última generació (estat de l'art) per a l'anàlisi del comportament de reactor. En el pla termohidràulic s'utilitza el codi acoblat desenvolupat, format per l'acoblament semi-implícit entre el codi de sis-tema TRACE i el de subcanal COBRA-TF (CTF), en una versió paral·lela creada en aquest treball. En transitoris en els que resulten necessaris els càlculs de neutrònica tridimensional, s'ha desenvolupat l'acoblament explícit entre el simulador tridimensio-nal de nuclis PARCS i el codi de subcanal CTF. Per a l'anàlisi de la integritat de les varetes de combustible s'empren els codis FRAPCON i FRAPTRAN, acoblant aquest últim de forma temporalment explícita amb CTF. Tots els desenvolupaments realitzats s'han inclòs en una mateixa plataforma informàti-ca que els engloba i coordina les simulacions sota les directrius de l'usuari. La plata-forma posseeix suficient flexibilitat per realitzar estudis de seguretat en multitud d'es-cenaris operacionals o accidentals, i es desitja que en un futur pugui ser utilitzada en càlculs de suport a llicència. Les ferramentes desenvolupades han sigut verificades mitjançant una sèrie d'aplicacions pràctiques en diferents transitoris i escenaris acci-dentals en reactors d'aigua lleugera. Els resultats obtinguts s'han comparat amb mesu-res reals de planta i amb els resultats obtinguts amb altres codis de simulació, mostrant una adequada capacitat predictiva. El treball realitzat en la present tesi doctoral s'emmarca dins de la línia d'investigació finançada pel Ministeri d'Economia i Competitivitat en el projecte NUC-MULTPHYS (ENE2012-34585) i els projectes de col·laboració interdisciplinar de la Universitat Politècnica de València COBRA_PAR (PAID-2810.11.05) i Open-NUC (PAID-05-12). / Abarca Giménez, A. (2017). Desarrollo y verificación de una plataforma multifísica de altas prestaciones para análisis de seguridad en ingeniería nuclear [Tesis doctoral no publicada]. Universitat Politècnica de València. https://doi.org/10.4995/Thesis/10251/88399