The Crash Modification Factor (CMF) clearinghouse can be used to estimate benefits for specific highway safety countermeasures. It assists safety professionals in the allocation of investments. The clearinghouse contains over 7000 entries of which only 446 are categorized as intelligent transportation systems or advanced technology, but none directly address connected or autonomous vehicles (CAVs). Further, the effectiveness of highway safety countermeasures is assumed to remain constant over time, an assumption that is particularly problematic as new technologies are introduced. For example, for the existing fleet of human-driven vehicles, installation of rumble strip can potentially reduce “run-off-road” crashes by 40%. If specific CAV technologies, e.g., lane-tracking, can work without rumble strips, and say, half of all cars are so equipped, only half of the fleet will benefit, reducing the benefits of rumble strips by a commensurate amount. Benefits of the two improvements, e.g., rumble strips and automated vehicles, should not be double-counted. As there will still be human-driven and/or non-connected vehicles in the fleet, conventional countermeasures are still necessary, although returns on conventional safety investments may be significantly overestimated. This is important as safety investments should be optimized and geared to future, not past fleets. Moreover, as CMFs are based on historical events, the types of crashes experienced by human-driven, un-connected cars are likely to be much different in the future. This research presents methods to estimate the safety benefits that autonomous vehicles have to offer and the changes needed in CMFs as a result of their adoption. This will primarily be achieved by modifying and enhancing a tool co-developed by the Fellow that estimates the safety benefits of different levels of autonomy. This tool, ddSAFCAT, estimates CAV safety benefits using real-world data for crashes, market penetration, and effectiveness.

This research focuses primarily on the analysis of exit ramp performance related to safety and operations. The safety analysis focuses on the impacts of different exit ramp types for freeway diverge areas and different factors contributing to the crashes that occur on the exit ramp sections. The operational analysis is based mainly on simulations by TSIS-CORSIM. Different ramp effects and guidance for selecting optimal exit ramp type are concluded. Issues related to ramp sections and crossroad sections are also demonstrated. Minimum ramp length and minimum distance between ramp terminal and downstream or upstream intersections are calculated. The operational analysis was conducted to determine different ramp effects and to provide guidance for selecting optimal exit ramp type. Comparisons of the operational performance of different types of exit ramps are made to present a method for choosing the optimal one. Some methods of evaluation (MOEs) are used to approach this objective, such as number of lane changes, average speed, delay time, etc. Data collection at 24 sites in Florida was conducted, and traffic simulations by TSIS-CORSIM were applied for analysis. Mathematical models were built to evaluate different impacts of these ramps based on simulations. All impact analysis is concluded to summarize a model for optimal exit ramp selection. In addition to ramp type evaluation and selection, issues related to ramp section and crossroad section are demonstrated. Minimum ramp length and minimum distance between ramp terminal and downstream or upstream intersections are calculated.

O licenciamento de uma instalação nuclear é motivado pela necessidade de proteger os seres humanos e o meio ambiente das radiações ionizantes e, ao mesmo tempo, define as bases para a concepção e a determinação da aceitabilidade da planta. Uma parte importante no processo de licenciamento é a realização de uma análise de acidentes, a qual deve estar documentada no Relatório Final de Análise de Segurança (RFAS). Existem diferentes opções de cálculo na área de acidentes, combinando a utilização de códigos computacionais e dados de entrada, para fins de licenciamento. Uma delas é a Best Estimate Plus Uncertainty (BEPU), que considera dados de entrada realistas e as incertezas associadas. As aplicações de abordagens BEPU em processos de licenciamento iniciaram-se nos anos 2000, primeiro para análise de Acidente de Perda de Refrigerante (Loss of Coolant Accident - LOCA), e depois para a análise de acidentes como um todo, documentados no Capítulo 15 do RFAS. O presente trabalho tem como objetivo principal demonstrar que é possível a aplicação da metodologia BEPU em todas as análises contidas no RFAS, identificando as disciplinas-chave do processo de licenciamento e os códigos computacionais utilizados. Este trabalho foi desenvolvido em conjunto com a Universidade de Pisa, Itália, com a colaboração do Prof. Dr. Francesco D\'Áuria. A principal motivação desse trabalho é o aprimoramento da metodologia BEPU para sua implementação em reatores do tipo PWR (Pressurized Water Reactor) no Brasil e no mundo, especialmente para fins de licenciamento, uma vez que as plantas nucleares brasileiras têm pouca experiência na área de cálculo de incertezas. / The licensing process of a nuclear power plant is motivated by the need to protect humans and the environment from ionizing radiation and, at the same time, sets out the basis for the design and determining the acceptability of the plant. An important part of the licensing process is the realization of accident analysis, which should be documented in the Final Safety Analysis Report (FSAR). There are different options on accidents calculation area by combining the use of computer codes and data entry for licensing purposes. One is the Best Estimate Plus Uncertainty (BEPU), which considers realistic input data and associated uncertainties. Applications of BEPU approaches in licensing procedures were initiated in the 2000s, first to analysis of Loss of Coolant Accident (LOCA), and then to the accident analysis as a whole, documented in Chapter 15 of the FSAR. This work has as main objective demonstrate the implementation of BEPU methodology in all analyses contained in FSAR is possible, identifying the key disciplines of the licensing process and the computer codes. This work was done in conjunction with the University of Pisa, Italy, with the collaboration of Professor Francesco D\'Auria. The main motivation of this work is the improvement of BEPU methodology for its implementation in PWR (Pressurized Water Reactor) reactors in Brazil and the world, especially for licensing purposes, since the Brazilian nuclear plants have little experience in the regulatory area, and specifically in calculation uncertainties.

O objetivo deste trabalho é apresentar a base de dados que foi desenvolvida para armazenar dados técnicos e processar dados sobre operação, falha e manutenção de equipamentos dos reatores nucleares de pesquisa localizados no Instituto de Pesquisas Energéticas e Nucleares (IPEN), em São Paulo - SP. Os dados extraídos desta base poderão ser aplicados na Análise Probabilística de Segurança dos reatores de pesquisa ou em avaliações quantitativas menos complexas relacionadas à segurança, confiabilidade, disponibilidade e manutenibilidade destas instalações. Esta base de dados foi desenvolvida de modo a permitir que as informações nela contidas estejam disponíveis aos usuários da rede corporativa, que é a intranet do IPEN. Os profissionais interessados deverão ser devidamente cadastrados pelo administrador do sistema, para que possam efetuar a consulta e/ou o manuseio dos dados. O modelo lógico e físico da base de dados foi representado por um diagrama de entidades e relacionamento e está de acordo com os módulos de segurança instalados na intranet do IPEN. O sistema de gerenciamento da base de dados foi desenvolvido com o MySQL, o qual utiliza a linguagem SQL como interface. A linguagem de programação PHP foi usada para permitir o manuseio da base de dados pelo usuário. Ao final deste trabalho, foi gerado um sistema de gerenciamento de base de dados capaz de fornecer as informações de modo otimizado e com bom desempenho. / The objective of this work is to present the computational database that was developed to store technical information and process data on component operation, failure and maintenance for the nuclear research reactors located at the Nuclear and Energy Research Institute (Instituto de Pesquisas Energéticas e Nucleares, IPEN), in São Paulo, Brazil. Data extracted from this database may be applied in the Probabilistic Safety Analysis of these research reactors or in less complex quantitative assessments related to safety, reliability, availability and maintainability of these facilities. This database may be accessed by users of the corporate network, named IPEN intranet. Professionals who require the access to the database must be duly registered by the system administrator, so that they will be able to consult and handle the information. The logical model adopted to represent the database structure is an entity-relationship model, which is in accordance with the protocols installed in IPEN intranet. The open-source relational database management system called MySQL, which is based on the Structured Query Language (SQL), was used in the development of this work. The PHP programming language was adopted to allow users to handle the database. Finally, the main result of this work was the creation a web application for the component reliability database named PSADB, specifically developed for the research reactors of IPEN; furthermore, the database management system provides relevant information efficiently.

O objetivo deste trabalho é apresentar a base de dados que foi desenvolvida para armazenar dados técnicos e processar dados sobre operação, falha e manutenção de equipamentos dos reatores nucleares de pesquisa localizados no Instituto de Pesquisas Energéticas e Nucleares (IPEN), em São Paulo - SP. Os dados extraídos desta base poderão ser aplicados na Análise Probabilística de Segurança dos reatores de pesquisa ou em avaliações quantitativas menos complexas relacionadas à segurança, confiabilidade, disponibilidade e manutenibilidade destas instalações. Esta base de dados foi desenvolvida de modo a permitir que as informações nela contidas estejam disponíveis aos usuários da rede corporativa, que é a intranet do IPEN. Os profissionais interessados deverão ser devidamente cadastrados pelo administrador do sistema, para que possam efetuar a consulta e/ou o manuseio dos dados. O modelo lógico e físico da base de dados foi representado por um diagrama de entidades e relacionamento e está de acordo com os módulos de segurança instalados na intranet do IPEN. O sistema de gerenciamento da base de dados foi desenvolvido com o MySQL, o qual utiliza a linguagem SQL como interface. A linguagem de programação PHP foi usada para permitir o manuseio da base de dados pelo usuário. Ao final deste trabalho, foi gerado um sistema de gerenciamento de base de dados capaz de fornecer as informações de modo otimizado e com bom desempenho. / The objective of this work is to present the computational database that was developed to store technical information and process data on component operation, failure and maintenance for the nuclear research reactors located at the Nuclear and Energy Research Institute (Instituto de Pesquisas Energéticas e Nucleares, IPEN), in São Paulo, Brazil. Data extracted from this database may be applied in the Probabilistic Safety Analysis of these research reactors or in less complex quantitative assessments related to safety, reliability, availability and maintainability of these facilities. This database may be accessed by users of the corporate network, named IPEN intranet. Professionals who require the access to the database must be duly registered by the system administrator, so that they will be able to consult and handle the information. The logical model adopted to represent the database structure is an entity-relationship model, which is in accordance with the protocols installed in IPEN intranet. The open-source relational database management system called MySQL, which is based on the Structured Query Language (SQL), was used in the development of this work. The PHP programming language was adopted to allow users to handle the database. Finally, the main result of this work was the creation a web application for the component reliability database named PSADB, specifically developed for the research reactors of IPEN; furthermore, the database management system provides relevant information efficiently.

This research focuses on the Crash Analysis Methodology for Segments (CAMS) which provides a way for engineers at the Utah Department of Transportation (UDOT) to prioritize safety improvements on state-owned roadways. Unlike the Utah crash analysis methodologies that come before it, the CAMS focuses exclusively on segment-related crashes. The benefits of such an analysis can be found in identifying locations that have safety concerns unbiased from intersections and their related crashes. The CAMS uses UDOT data to create a spreadsheet of roadway segments and their associated crashes. Each segment is homogeneous with respect to five variables: Annual Average Daily Traffic (AADT), functional class, number of lanes, speed limit, and urban code. In the statistical analyses performed on the data, four years of crash data (2014-2017) are used to predict distributions of crashes for the most recent year of data (2018). Observed crash counts are compared to the predicted distributions and assigned a percentile value within the distributions, and segments are subsequently ranked in order of safety concern according to those percentiles. Two-page technical reports are created for segments that rank high in the state or UDOT Region. These reports consist of concise tables of roadway data and crash trends pertaining to each segment. Research analysts also add observations made in virtual site visits to the reports. In the end, the results and the reports are sent to UDOT where UDOT Region engineers may review and study identified segments in further detail. This research also includes modifications made to the Intersection Safety Analysis Methodology (ISAM) which focuses exclusively on intersection-related crashes. The modifications made to the ISAM mirror the abilities of the CAMS, thus allowing the pair of methodologies to analyze the entire state route network without overlapping any crash data.

Die evangelische St. Pankratius-Kirche in Lucka wird aktuell genutzt und dient wöchentlich zehn bis 20 Menschen als Ort der Messe. Die Kirche befindet sich zu Teilen in einem sanierungsbedürftigen Zustand, vor allem im Dachstuhl sind Schäden an der historischen Bausubstanz nicht zu übersehen. Zum Erhalt dieses Baudenkmals ist neben der Sanierung und konstruktiven Ertüchtigung auch ein Konzept zur erweiterten Nutzung geplant. In Anbetracht sinkender Mitgliederzahlen der Kirche ist eine Erweiterung des kulturellen Angebotes in diesem Gotteshaus notwendig, um die aktive Nutzung auch in den nächsten Jahrzehnten sicherzustellen. Diese Thesis soll als Vorarbeit und Grundlage für die Umsetzung von Sanierungsmaßnahmen dienen. Darin soll zum Einen die Tragkonstruktion des Daches vermessen und aus statischer Sicht betrachtet werden. Zum Anderen erfolgt eine Bauzustandsanalyse, welche Schadensuntersuchungen und Sanierungsempfehlungen beinhaltet.:1. Einleitung 1.1. Motivation 1.2. Aufbau der Arbeit 2. Vorstellung des Objektes 2.1. Geschichtliche Einordnung 2.2. Beschreibung der Kirche 3. Vermessung und Aufnahme des Dachstuhls 3.1. Messtechnik und Messablauf 3.2. Bestandsaufnahme 3.2.1. Bauteile 3.2.2. Tragwerk 3.2.3. Verbindungen 3.2.4. Bauteilquerschnitte 3.2.5. Visuelle Sortierung und Festigkeit 3.2.6. Verformungen 3.3. Überführung in ein 3D-Modell 4. Holz als Baustoff 4.1. Allgemein 4.2. Bemessung von Holz 4.3. Versagensmechanismen im Holzbau 4.4. Materialkennwerte von Holz 5. Lastannahme nach DIN EN 1991-1 5.1. Eigenlasten 5.2. Nutzlasten 5.3. Schneelasten 5.4. Windlasten 5.5. Zusammenfassung Lastfälle 6. Bildung des statischen Modells 6.1. Kraftfluss 6.2. Modellierung der Verbindungen 6.3. Tragverhalten 6.4. Statische Systeme 7. Statische Nachweise des Ist-Zustandes 7.1. Szenario 1 - System intakt 7.2. Szenario 2 - Ausfall Sparrenfußpunkt 8. Erfassen der Schadensbilder 8.1. Untersuchungen 8.1.1. Visuelle Untersuchungen 8.1.2. Feuchtemessung 8.1.3. Nachweis Sulfat und Bor 8.1.4. Nachweis chlorierter Kohlenwasserstoffe 8.1.5. Bestimmung der Käfernagsel 8.2. Bestimmung der Holzarten 8.3. Fotodokumentation 9. Schadensursachen 9.1. Biologische Einflüsse 9.1.1. Pilzbefall 9.1.2. Insektenbefall 3 9.2. Mechanische Einflüsse 9.3. Bauphysikalische Einflüsse 10.Bauliche Maßnahmen 10.1. Konstruktiver Holzschutz 10.2. Chemischer Holzschutz 10.3. Instandsetzung und Ertüchtigung 11.Denkmalschutz 12.Wirtschaftlichkeitsanalyse 13.Sanierungskonzept 14.Statische Nachweise der Ertüchtigungen 15.Zusammenfassung / The protestant St. Pankratius church in Lucka is currently in use and serves ten to 20 people as a place of mass every week. Parts of the church require redevelopment, especially the roof truss showing damage to the historic structure. To preserve this architectural monument, in addition to the renovation and structural reinforcement, a concept for expanded use is foreseen. Given the church’s declining membership, expanding the cultural offerings in this house of worship is necessary to ensure its active use in the coming decades. This thesis aims to serve as preliminary work and a basis for implementing renovation measures. It consists of two parts: First, the roof’s supporting structure will be measured and examined from a structural point of view. Second, the condition of the building will be analysed, including an investigation of damages and recommendations for renovations.:1. Einleitung 1.1. Motivation 1.2. Aufbau der Arbeit 2. Vorstellung des Objektes 2.1. Geschichtliche Einordnung 2.2. Beschreibung der Kirche 3. Vermessung und Aufnahme des Dachstuhls 3.1. Messtechnik und Messablauf 3.2. Bestandsaufnahme 3.2.1. Bauteile 3.2.2. Tragwerk 3.2.3. Verbindungen 3.2.4. Bauteilquerschnitte 3.2.5. Visuelle Sortierung und Festigkeit 3.2.6. Verformungen 3.3. Überführung in ein 3D-Modell 4. Holz als Baustoff 4.1. Allgemein 4.2. Bemessung von Holz 4.3. Versagensmechanismen im Holzbau 4.4. Materialkennwerte von Holz 5. Lastannahme nach DIN EN 1991-1 5.1. Eigenlasten 5.2. Nutzlasten 5.3. Schneelasten 5.4. Windlasten 5.5. Zusammenfassung Lastfälle 6. Bildung des statischen Modells 6.1. Kraftfluss 6.2. Modellierung der Verbindungen 6.3. Tragverhalten 6.4. Statische Systeme 7. Statische Nachweise des Ist-Zustandes 7.1. Szenario 1 - System intakt 7.2. Szenario 2 - Ausfall Sparrenfußpunkt 8. Erfassen der Schadensbilder 8.1. Untersuchungen 8.1.1. Visuelle Untersuchungen 8.1.2. Feuchtemessung 8.1.3. Nachweis Sulfat und Bor 8.1.4. Nachweis chlorierter Kohlenwasserstoffe 8.1.5. Bestimmung der Käfernagsel 8.2. Bestimmung der Holzarten 8.3. Fotodokumentation 9. Schadensursachen 9.1. Biologische Einflüsse 9.1.1. Pilzbefall 9.1.2. Insektenbefall 3 9.2. Mechanische Einflüsse 9.3. Bauphysikalische Einflüsse 10.Bauliche Maßnahmen 10.1. Konstruktiver Holzschutz 10.2. Chemischer Holzschutz 10.3. Instandsetzung und Ertüchtigung 11.Denkmalschutz 12.Wirtschaftlichkeitsanalyse 13.Sanierungskonzept 14.Statische Nachweise der Ertüchtigungen 15.Zusammenfassung

De nos jours, le degré de complexité des systèmes de transport notamment ferroviaires est important et grandissant. En effet, les nouvelles technologies et besoins des grandes métropoles contraignent les concepteurs de ces systèmes à toujours intégrer de plus en plus de fonctionnalités. Cette complexité rend obsolète les méthodes de conception traditionnelles qui reposent sur l'expertise d'ingénieurs. La conception du système est définie sur plusieurs documents de spécification, de centaines de pages chacun. Ces derniers comprennent généralement la description des différentes exigences, fonctionnalités et sous-systèmes, ils sont rédigés en langage naturel. Ceci est source d'erreurs et ambigüités lors des différentes phases de conception et de validation sécuritaire. De ce fait, l'ingénierie système basée sur les modèles propose des solutions outillées afin de structurer et spécifier de manière systématique le fonctionnement du système. Dans cette thèse, nous proposons une approche de rétro-ingénierie de ces spécifications système. Cette approche nous permet de définir une méthodologie orientée-scénarios pour la formalisation et analyse de spécifications système. Elle est basée sur le langage de modélisation semi-formel ScOLa (Scénarios Oriented Language) et un modèle d'exécution formel basé sur le langage AltaRica 3.0. Nous proposons, á partir de ScOLa, des techniques contribuant aux analyses de sûreté fonctionnelles du système. Ces techniques permettent de clarifier, structurer et analyser / As of today, the complexity of transportation systems and especially railways is continuously growing. Indeed, new technologies and needs of major metropolitan areas constrain system designers to integrate more and more functionalities. Traditional methods for the design of systems are mainly based on the knowledge of experts. Systems are specified in several documents, of hundreds of pages each. These documents provide a general description of the requirements, functionalities and sub-systems, and are written in a natural language. This is error-prone and source of ambiguities. Model-based systems engineering proposes systematic solutions to structure and specify systems architectures. In this thesis, we propose a reverse engineering approach that is able to capture systems concepts. This leads us to propose ScOLa, a scenario-oriented modeling language for systems specifications with the following features:- A scenario-oriented approach to provide a more contextual view of the system, at several abstraction levels.- A co-construction of the architecture and behavior of the system in the same model.- The use of ScOLa for a seamless integration of system specification and safety analysis. We propose ascenario-oriented methodology for safety analysis, implemented using AltaRica 3.0.

There is so far no consensus on how to develop a reliability model of safety-related digitalinstrumentation and control (I&C) in a probabilistic safety assessment (PSA) of a nuclear powerplant. The objective of this thesis is to evaluate different approaches to model digital protectionsystems in a PSA for a nuclear power plant. This is accomplished by the development of a fault treemodel of the digital protection system for a fictive and simplified nuclear power plant, that act as areference model to be used for evaluation of different design alternatives and modelling principles.Common cause failures and spurious actuation signals are the major contributors to scenariosresulting in a core damage. A PSA model has to be sufficiently detailed in order for this to berepresented in the results. The impact on results such as core damage frequency and importance ofminimal cut sets from different fail-safe, voting logic and signal validation designs are significant,too. To further examine the differences between I&C designs and significance of different PSAmodelling solutions, the degree of realism of the example should be increased. This rapidlyintroduces complexity to the models resulting in a model that is more difficult to review and resultsthat is more difficult to interpret and even much simplified models tend to get rather complex.

McMaster University DOCTOR OF PHILOSOPHY (2022) Hamilton, Ontario (Engineering Physics) TITLE: Development and Evaluation of Polaris CANDU Geometry Modelling and of TRACE_Mac/PARCS_Mac Coupling with RRS for CANDU Analysis AUTHOR: Simon Younan, M.A.Sc. (McMaster University), B.Eng. (McMaster University) SUPERVISOR: Dr. David Novog NUMBER OF PAGES: xiv, 163 / In the field of nuclear safety analysis, as computers have become more powerful, there has been a trend away from low-fidelity models using conservative assumptions, to high-fidelity best-estimate models combined with uncertainty analysis. A number of these tools have been developed in the United States, due to the popularity of light water reactors. These include the SCALE analysis suite developed by ORNL, as well as the PARCS and TRACE tools backed by the USNRC. This work explores adapting the capabilities of these tools to the analysis of CANDU reactors. The Polaris sequence, introduced in SCALE 6.2, was extended in this work to support CANDU geometries and compared to existing SCALE sequences such as TRITON. Emphasis was placed on the Embedded Self-Shielding Method (ESSM), introduced with Polaris. Both Polaris and ESSM were evaluated and found to perform adequately for CANDU geometries. The accuracy of ESSM was found to improve when the precomputed selfshielding factors were updated using a CANDU representation. The PARCS diffusion code and the TRACE system thermalhydraulics code were coupled, using the built-in coupling capability between the two codes. In addition, the Exterior Communications Interface (ECI), used for coupling with TRACE, was utilized. A Python interface to the ECI library was developed in this work and used to couple an RRS model written in Python to the coupled PARCS/TRACE model. A number of code modifications were made to accommodate the required coupling and correct code deficiencies, with the modified versions named PARCS_Mac and TRACE_Mac. The coupled codes were able to simulate multiple transients based on prior studies as well as operational events. The code updates performed in this work may be used for many future studies, particularly for uncertainty propagation through a full set of calculations, from the lattice model to a full coupled system model. / Thesis / Doctor of Philosophy (PhD) / Modern nuclear safety analysis tools offer more accurate predictions for the safety and operation of nuclear reactors, including CANDU reactors. These codes take advantage of modern computer hardware, and also a shift in philosophy from conservative analysis to best estimate plus uncertainty analysis. The goal of this thesis was to adapt a number of modern tools to support CANDU analysis and uncertainty propagation, with a particular emphasis on coupling of multiple interacting models. These tools were then demonstrated, and results analyzed. The simulations performed in this work were successful in producing results comparable to prior studies along with experimental and operational data. This included the simulation of four weeks of reactor operation including “shim mode” operation. Sensitivity and uncertainty analyses were performed over the course of the work to quantify the precision and significance of the results as well as to identify areas of interest for future research.