Les Etudes Probabilistes de Sûreté (EPS) de niveau 2 en centrale nucléaire visent à identifier les séquences d’événements pouvant correspondre à la propagation d’un accident d’un endommagement du cœur jusqu’à une perte potentielle de l’intégrité de l’enceinte, et à estimer la fréquence d’apparition des différents scénarios possibles.<p>Ces accidents sévères dépendent non seulement de défaillances matérielles ou d’erreurs humaines, mais également de l’occurrence de phénomènes physiques, tels que des explosions vapeur ou hydrogène. La prise en compte de tels phénomènes dans le cadre booléen des arbres d’événements s’avère difficile, et les méthodologies dynamiques de réalisation des EPS sont censées fournir une manière plus cohérente d’intégrer l’évolution du processus physique dans les changements de configuration discrète de la centrale au long d’un transitoire accidentel.<p>Cette thèse décrit l’application d’une des plus récentes approches dynamiques des EPS – la Théorie de la Dynamique Probabiliste basée sur les Stimuli (SDTPD) – à différents modèles de déflagration d'hydrogène ainsi que les développements qui ont permis cette applications et les diverses améliorations et techniques qui ont été mises en oeuvre.<p><p>Level-2 Probabilistic Safety Analyses (PSA) of nuclear power plants aims to identify the possible sequences of events corresponding to an accident propagation from a core damage to a potential loss of integrity of the containment, and to assess the frequency of occurrence of the different scenarios.<p>These so-called severe accidents depend not only on hardware failures and human errors, but also on the occurrence of physical phenomena such as e.g. steam or hydrogen explosions. Handling these phenomena in the classical Boolean framework of event trees is not convenient, and dynamic methodologies to perform PSA studies are expected to provide a more consistent way of integrating the physical process evolution with the discrete changes of plant configuration along an accidental transient.<p>This PhD Thesis presents the application of one of the most recently proposed dynamic PSA methodologies, i.e. the Stimulus-Driven Theory of Probabilistic Dynamics (SDTPD), to several models of hydrogen explosion in the containment of a plant, as well as the developed methods and improvements.<p> / Doctorat en Sciences de l'ingénieur / info:eu-repo/semantics/nonPublished

Current regulations specify the minimum number of operators required per nuclear power plant. However, these requirements are based on the operation of large nuclear power plants, which are not inherent safe and can result in a meltdown. For newly developed small nuclear reactors, the current number of operators seems to be excessive causing the technology to be less competitive. Before the number of required operators can be optimised, it should be demonstrated that human errors will not endanger or cause risk to the plant or public. For this study, a small pebble bed High Temperature Reactor (HTR) Nuclear Power Plant (NPP), the Th-100, was evaluated. The inherent safety features of this type of nuclear reactor include independent barriers for fission product capture and passive heat dissipation during a loss of coolant. The control and instrumentation architecture include two independent protection systems. The Control and Limitation System is the first protection system to react if the reactor parameters exceed those of the normal operational safe zone. If the Control and Limitation System fail to maintain the reactor within the safe zone, the Reactor Protection System would at that time operate and force the reactor to a safe state. Both these automated protection systems are installed in a control room local to the reactor building, protected from adverse conditions. In addition, it is connected to a semi-remote control room, anticipated as a multi-unit control room to include the monitoring and control of the auxiliary systems. Probable case studies of human error associated with multi-unit control rooms were evaluated against the logic of the Control and Limitation System. Fault Tree Analysis was used to investigate all possible failures. The evaluation determined the reliability of the Control and Limitation System and highlighted areas which design engineers should take into account if a higher reliability is required. The scenario was expanded, applying the same methods, to include the large release of fission products in order to verify the reliability calculations. The probability of a large release of fission products compared with studies done on other nuclear installations revealed to be much less for the evaluated HTR as was expected. As the study has proved that human error cannot have a negative influence on the safety of the reactor, it can be concluded that the first step has been met which is required, when applying for a waiver to utilise a multi-unit control room for the small pebble bed HTR NPP. Also, from the study, it is recommended that a practical approach be applied for the evaluation of operator duties on a live plant, to optimise the number of operators required. This in turn will position the inherent safe HTR competitively over other power stations. / MIng (Nuclear Engineering), North-West University, Potchefstroom Campus, 2015

Current regulations specify the minimum number of operators required per nuclear power plant. However, these requirements are based on the operation of large nuclear power plants, which are not inherent safe and can result in a meltdown. For newly developed small nuclear reactors, the current number of operators seems to be excessive causing the technology to be less competitive. Before the number of required operators can be optimised, it should be demonstrated that human errors will not endanger or cause risk to the plant or public. For this study, a small pebble bed High Temperature Reactor (HTR) Nuclear Power Plant (NPP), the Th-100, was evaluated. The inherent safety features of this type of nuclear reactor include independent barriers for fission product capture and passive heat dissipation during a loss of coolant. The control and instrumentation architecture include two independent protection systems. The Control and Limitation System is the first protection system to react if the reactor parameters exceed those of the normal operational safe zone. If the Control and Limitation System fail to maintain the reactor within the safe zone, the Reactor Protection System would at that time operate and force the reactor to a safe state. Both these automated protection systems are installed in a control room local to the reactor building, protected from adverse conditions. In addition, it is connected to a semi-remote control room, anticipated as a multi-unit control room to include the monitoring and control of the auxiliary systems. Probable case studies of human error associated with multi-unit control rooms were evaluated against the logic of the Control and Limitation System. Fault Tree Analysis was used to investigate all possible failures. The evaluation determined the reliability of the Control and Limitation System and highlighted areas which design engineers should take into account if a higher reliability is required. The scenario was expanded, applying the same methods, to include the large release of fission products in order to verify the reliability calculations. The probability of a large release of fission products compared with studies done on other nuclear installations revealed to be much less for the evaluated HTR as was expected. As the study has proved that human error cannot have a negative influence on the safety of the reactor, it can be concluded that the first step has been met which is required, when applying for a waiver to utilise a multi-unit control room for the small pebble bed HTR NPP. Also, from the study, it is recommended that a practical approach be applied for the evaluation of operator duties on a live plant, to optimise the number of operators required. This in turn will position the inherent safe HTR competitively over other power stations. / MIng (Nuclear Engineering), North-West University, Potchefstroom Campus, 2015

Thesis (MBA)--Stellenbosch University, 2002. / ENGLISH ABSTRACT: The Pebble Bed Modular Reactor is a new and inherently safe concept for a nuclear power generation plant. In order to obtain the necessary licenses to build and operate this reactor, numerous design and safety analyses need to be performed. The results of these analyses must be supported with substantial proof to provide the nuclear authorities with a sufficient level of confidence in these results to be able to supply the required licences. Beside the obvious need for a sufficient level of confidence in the safety analyses, the analyses concerned with investment protection also need to be reliable from the investors’ point of view. The process to be followed to provide confidence in these analyses is the verification and validation process. It is aimed at presenting reliable material against which to compare the results from the simulations. This material for comparison will consist of a combination of results from experimental data, extracts from actual plant data, analytical solutions and independently developed solutions for the simulation of the event to be analysed. Besides comparison with these alternative sources of information, confidence in the results will also be built by providing validated statements on the accuracy of the results and the boundary conditions with which the simulations need to comply. Numerous standards exist that address the verification and validation of computer software, for instance by organisations such as the American Society of Mechanical Engineers (ASME) and the Institute of Electrical and Electronics Engineers (IEEE). The focal points of the verification and validation of the design and safety analyses performed on typical PBMR modes and states, and the requirements imposed by both the local and overseas nuclear regulators, are not entirely enveloped by these standards. For this reason, PBMR developed a systematic and disciplined approach for the preparation of the Verification and Validation Plan, aimed at capturing the essence of the analyses. This approach aims to make a definite division between software development and the development of technical analyses, while still using similar processes for the verification and validation. The reasoning behind this is that technical analyses are performed by engineers and scientists who should only be responsible for the verification and validation of the models and data they use, but not for the software they are dependent on. Software engineers should be concerned with the delivery of qualified software to be used in the technical analyses. The PBMR verification and validation process is applicable to both hand calculations and computer-aided analyses, addressing specific requirements in clearly defined stages of the software and Technical Analysis life cycle. The verification and validation effort of the Technical Analysis activity is divided into the verification and validation of models and data, the review of calculational tasks, and the verification and validation of software, with the applicable information to be validated, captured in registers or databases. The resulting processes are as simple as possible, concise and practical. Effective use of resources is ensured and internationally accepted standards have been incorporated, aiding in faith in the process by all stakeholders, including investors, nuclear regulators and the public. / AFRIKAASE OPSOMMING: Die Modulêre Korrelbedreaktor is ’n nuwe konsep vir ’n kernkragsentrale wat inherent veilig is. Dit word deur PBMR (Edms.) Bpk. ontwikkel. Om die nodige vergunnings om so ’n reaktor te kan bou en bedryf, te bekom, moet ’n aansienlike hoeveelheid ontwerp- en veiligheidsondersoeke gedoen word. Die resultate wat hierdie ondersoeke oplewer, moet deur onweerlegbare bewyse ondersteun word om vir die owerhede ’n voldoende vlak van vertroue in die resultate te gee, sodat hulle die nodigde vergunnings kan maak. Benewens die ooglopende noodsaak om ’n voldoende vlak van vertroue in die resultate van die veiligheidsondersoeke te hê, moet die ondersoeke wat met die beskerming van die beleggers se beleggings gepaard gaan, net so betroubaar wees. Die proses wat gevolg word om vertroue in die resultate van die ondersoeke op te bou, is die proses van verifikasie en validasie. Dié proses is daarop gerig om betroubare vergelykingsmateriaal vir simulasies voor te lê. Hierdie vergelykingsmateriaal vir die gebeurtenis wat ondersoek word, sal bestaan uit enige kombinasie van inligting wat in toetsopstellings bekom is, wat in bestaande installasies gemeet is, wat analities bereken is; asook dit wat deur ’n derde party onafhanklik van die oorspronklike ontwikkelaars bekom is. Vertroue in die resultate van die ondersoeke sal, behalwe deur vergelyking met hierdie alternatiewe bronne van inligting, ook opgebou word deur die resultate te voorsien van ’n gevalideerde verklaring wat die akkuraatheid van die resultate aantoon en wat die grensvoorwaardes waaraan die simulasies ook moet voldoen, opsom. Daar bestaan ’n aansienlike hoeveelheid internasionaal aanvaarde standaarde wat die verifikasie en validasie van rekenaarsagteware aanspreek. Die standaarde kom van instansies soos die Amerikaanse Vereniging vir Meganiese Ingenieurs (ASME) en die Instituut vir Elektriese en Elektroniese Ingenieurs (IEEE) – ook van Amerika. Die aandag wat deur die Suid-Afrikaanse en oorsese kernkragreguleerders vereis word vir die toestande wat spesifiek geld vir korrelbedreaktors, word egter nie geheel en al deur daardie standaarde aangespreek nie. Daarom het die PBMR maatskappy ’n stelselmatige benadering ontwikkel om verifikasie- en validasieplanne voor te berei wat die essensie van die ondersoeke kan ondervang. Hierdie benadering is daarop gemik om ’n duidelike onderskeid te maak tussen die ontwikkeling van sagteware en die ontwikkeling van tegniese ondersoeke, terwyl steeds gelyksoortige prosesse in die verifikasie en validasie gebruik sal word. Die rede hiervoor is dat tegniese ondersoeke uitgevoer word deur ingenieurs en wetenskaplikes wat net vir verifikasie en validasie van hulle eie modelle en die gegewens verantwoordelik gehou kan word, maar nie vir die verifikasie en validasie van die sagteware wat hulle gebruik nie. Ingenieurs wat spesialiseer in sagteware-ontwikkeling behoort verantwoordelik te wees vir die daarstelling van sagteware wat deur die reguleerders gekwalifiseer kan word, sodat dit in tegniese ondersoeke op veiligheidsgebied gebruik kan word. Die verifikasie- en validasieproses van die PBMR is sowel vir handberekeninge as vir rekenaarondersteunde-ondersoek geskik. Hierdie proses spreek spesifieke vereistes in onderskeie stadiums gedurende die lewenssiklusse van die ontwikkeling van sagteware en van tegniese ondersoeke aan. Die verifikasie- en validasiewerk vir tegniese ondersoeksaktiwiteite is verdeel in die verifikasie en validasie van modelle en gegewens, die nasien van berekeninge en die verifikasie en validasie van sagteware, waarby die betrokke inligting wat gevalideer moet word, versamel word in registers of databasisse. Die prosesse wat hieruit voortgevloei het, is so eenvoudig as moontlik, beknop en prakties gehou. Hierdeur is ’n effektiewe benutting van bronne verseker. Internasionaal aanvaarde standaarde is gebruik wat die vertroue in die proses deur alle betrokkenes, insluitende beleggers, die owerhede en die publiek, sal bevorder.

本文共乙冊，約八萬言，分為五章十節。 第壹章 緒論。首先闡釋本文題目之意義；其次陳述本文之研究動機、目的方法與限 制；最後說明本文研究之架構。 第貳章 假設分析的運作背景。從組織環境與組織問題的分析中，以瞭解運用假設分 析方法之緣由。 第參章 假設分析的運作。首先論述假設分析運作的理論基礎；其次解析其運作之程 序與方法。 第肆章 假設分析的應用。首先說明政策論證的內涵，模式及其與假設分析方法之關 係；其次試圖運用本文所論述之理論與方法，以檢視和評析台灣電力公司興建核能四 廠一案的政策過程。 第伍章 結論。發表本文研究之心得並提出假設分析方法成功運作的要件。

本文共壹冊，分柒章，約四萬言，章節目錄如下： 第一章：導論 一、前言，二、研究動機，三、研究目的，四、研究架構，五、研究限制。 第二章：緊急事故本質探討 一、名詞解釋，二、緊急事故性質，三、緊急事故影響與後果，四、面臨緊急事故時 之個人與組織行為，五、災變之防治。 第三章：緊急事故的管理 一、管理架構，二、管理之規劃與控制活動。 第四章：決策支援系統理論基礎 一、決策支援系統定義與特性，二、傳統EDP,MIS 與DSS 之比較，三、系統建立方法 。 第五章：緊急事故管理之決策支援系統設計 一、緊急事故下之決策程序，二、決策特徵，三、功能架構，四、系統建立。 第六章：個案實例 一、個案背景介紹，二、核能電廠緊急應變措施，三、系統需求，四、建立與實施。 第七章：結論與建議 一、結論，二、建議。 環繞人類四周環境中，常有許多不確定的災變隨時可能降臨。且發生災變時，如果資 訊缺乏或運用不當，常造成不必要的損失與傷亡。本研究係研究有關決策支援系統在 緊急事故管理上之應用，籍資訊的提供與利用，以支援緊急事故防治。 決策支援系統具有（一）易於使用，（二）模式庫與資料庫整合，（三）適於解決非 結構性問題，（四）具有良好彈性等特性，而緊急事故管理更明顯涉及決策者的價值 判斷，及對不確定環境的偏好，因此面臨這種結構程度低的問題，一個考慮完整的決 策支援系統將能提供管理人員更有效的支援。 本文之研究共分四部份，第一部份探討緊急事故的特性與本質，及面臨災變時人與群 體的行為與反應，第二部份探討緊急事故管理之決策程序與決策特性，第三部份則以 文獻分析方式探討決策支援系統相關文獻，以作為建造系統時指引，第四部份則以個 案研究法，針對核能電廠緊急事故之疏散掩蔽決策，提出應建立之決策支援系統。

This thesis examines the opposition to civil nuclear energy in France and West Germany during the 1970s, arguing that small-scale interactions among its diverse participants led to broad changes in their personal lives and political environments. Drawing extensively on oral history interviews with former activists as well as police reports, media coverage and protest ephemera, this thesis shows how individuals at the grassroots built up a movement that transcended national (and social) borders. They were able to do so in part because nuclear power was such a multivalent symbol at the time. Residents of towns near planned power stations felt that nuclear technology represented an intervention in their community by state and industry, a potential threat to their health, wealth and way of life. In the decade after 1968, concerns like these coalesced with criticisms of capitalism, the state, militarism and consumer society that were being made by a more politicised constituency. This made the anti-nuclear movement both broad-based and highly fragmented. Activist networks linked people across existing national, political and social boundaries, but the social world of activism was subject to its own divisions (such as between locals and outsiders or between militant and non-violent activists). By analysing both the transnational dimensions and internal divisions of the anti-nuclear movement, this thesis revises the homogenising concepts of social movements that are prevalent in much of the existing sociological and political science literature. At the same time, it situates the anti-nuclear movement historically within the decade of upheaval that was the 1970s, while moving individual activists from the margins to the centre of protest history.

A Agência Internacional de Energia Atômica (IAEA - \"International Atomic Energy Agency\") vem incentivando o desenvolvimento de sistemas passivos de refrigeração em plantas nucleares visando a simplificação e o incremento da confiabilidade em funções essenciais de segurança nos projetos de uma próxima geração de reatores nucleares refrigerados a água. O principal fundamento desses sistemas é o emprego da circulação natural como sistema de segurança aplicável em operações de desligamento do reator para manutenção ou na ocorrência de acidentes. A circulação natural é um fenômeno que surge em virtude do gradiente de temperatura em pontos diferentes do circuito de refrigeração. Em condições extremas de estabilidade têm-se o estabelecimento do escoamento bifásico gás/líquido podendo configurar-se segundo diferentes regimes. A fração de vazio é reconhecida como um dos parâmetros chave na predição da ocorrência de instabilidades do escoamento bifásico. Apresenta-se neste trabalho uma inovadora metodologia para estimativa da fração de vazio a partir de imagens digitais capturadas diretamente de circuitos experimentais que geram o escoamento bifásico. O método é baseado na aquisição de imagens, com controle da profundidade de campo, de uma seção do Circuito de Circulação Natural (CCN) presente no IPEN/CNEN-SP. A imagem é segmentada com base na inferência fuzzy de diferentes parâmetros de segmentação e ajustada ao foco utilizado na sua aquisição. Ela é varrida de um modo inédito e iterativo, utilizando máscaras de diferentes tamanhos integrando um conjunto de redes neurais com a Transformada Randomizada de Hough. Cada diferente tamanho de máscara é escolhido de acordo com os tamanhos das bolhas que são os objetos de interesse. O volume da bolha é estimado baseado em sua projeção plana capturada nas imagens digitais. O cálculo da fração de vazio considera o volume da seção geométrica do escoamento no tubo de vidro cilíndrico e a profundidade de campo utilizada e nos parâmetros geométricos inferidos para cada bolha detectada. Os resultados mostraram que a integração entre o conjunto de redes neurais e a Transformada Randomizada de Hough aumentaram a robustez das estimativas do sistema. / The International Atomic Energy Agency (IAEA) has been encouraging the use of passive cooling systems in new designs of nuclear power plants. Next nuclear reactor generations are intended to possess simpler and robust safety functions. Natural circulation based systems hold an undoubtedly prominent position among these. Natural circulation phenomenon occurrence depends only on the existence of refrigerant liquid temperature gradient in different sections of the plant refrigerator circuit. The study of limit conditions for these systems has led to instability behavior analysis where many different two-phase flow patterns are present. Void fraction is a key parameter in thermal transfer analysis of theses flow instability conditions. This works presents a new method to estimate void fraction from digital images captured at an experimental two-phase flow circuit. The method is primarily based on depth-of-field controlled image acquisition of a section of a closed loop of natural circulating water through cylindrical glass tubes. This loop is called Natural Circulation Facility (NCF) and is located at Nuclear and Engineering Research Institute in Brazil (IPEN/CNEN-SP). Image is segmented based on fuzzy inference of different segmentation parameters and adjusted to image acquisition focus. The image is then scanned in an inedited way using different-sized masks integrating a set of different artificial neural networks with a modified Randomized Hough Transform. Each different mask size is chosen in accordance to bubble sizes which are objects of interest. The bubble volume is estimated based on two-dimensional projection sizing based on digitally acquired images. Void fraction calculation takes into account the volume of the geometrical section of flow inside cylindrical glass tube considering used depth-of-field. It is also based on the summed bubble geometrical parameters inferred for each detected bubble. The results have shown that integration between artificial-neural-net sets and Randomized Hough Transforms increase system estimations robustness.

L’accident nucléaire de Fukushima en mars 2011 a eu des répercussions politiques différentes dans la trentaine d’États exploitant un parc de centrales. Tandis que l’Allemagne décidait d’accélérer la sortie du nucléaire amorcée dix ans auparavant, la Suède abrogeait le moratoire introduit en 1981 sur la construction de nouveaux réacteurs et la France ne s’engageait qu’à diminuer marginalement la part du nucléaire dans son mix électrique. Trois pays européens, confrontés à un même évènement, prenaient ainsi trois directions opposées. Par le passé, les différences de politiques nucléaires ont été expliquées par des déterminismes géographiques, culturels, évènementiels ou économiques. Cette thèse propose de dépasser ces approches pour considérer ces choix comme le résultat de rapports de forces entre opposants et soutiens à l’énergie atomique s’affrontant, à plusieurs échelles, pour contrôler l’usage du territoire. En s’appuyant sur la méthode de la géopolitique locale, cette recherche vise à mettre en évidence les rivalités de pouvoirs et les représentations qui structurent les systèmes d’acteurs dans chacun des pays, ainsi que les stratégies mises en œuvre. Dans un contexte de transition énergétique où le nucléaire est présenté comme une solution aux bouleversements climatiques, il s’agit ici, grâce à une approche comparative, d’identifier en Allemagne, en France et en Suède, les éléments ayant conduit à l’élaboration de politiques énergétiques diamétralement différentes. / The Fukushima atomic disaster had different political fallouts in the thirty-one countries where nuclear power is exploited. In Europe, while Germany decided to accelerate the phase-out engaged ten years before, the Swedish government repealed the moratorium on new nuclear reactors introduced in 1981 and France only committed to reduce marginally the share of nuclear electricity. Three European countries, facing the same event, took three different directions. In the past, differences between countries’ nuclear policies have been explained by economic, geographical or cultural determinism. This work offers to go beyond these approaches and to consider energy policies as the result of power struggles between opponents and supporters of atomic energy who fight to control the territory. Thanks to the local geopolitical approach, this thesis aims at highlighting the rivalries and the representation which structure the actors’ systems in each country as well as the strategies used in the conflict. In a context where nuclear energy is portrayed as a potential solution to mitigate climate change, the purpose of this work is to identity the elements which led to the elaboration of diametrically opposed energy policies in France, Germany and Sweden.

Nenhuma / No circuito primário de uma usina nuclear do tipo PWR (Pressurized Water Reactor), o refrigerante do reator é mantido a uma temperatura interna por volta de 300 C e pressão interna da ordem de 15,0 MPa, durante operação normal. O Vaso de Pressão do Reator (VPR) contém os elementos combustíveis e é considerado o componente mais importante do circuito primário. A integridade do VPR deve ser assegurada durante toda a vida útil da usina, de forma a proteger os trabalhadores da usina e o público em geral dos danos decorrentes da liberação de material radioativo.Uma das condições de carregamento mais severas que pode ameçar a integridade do VPR é causada por um transitório conhecido como Choque Térmico Pressurizado (PTS - Pressurized Thermal Shock). O VPR estará sujeito a tal condição durante um acidente com perda de refrigerante do núcleo do reator. Em um evento como este, o sistema de refrigeração de emergência do núcleo é ativado, o que provoca a injeção de água fria no interior do VPR e, consequentemente, um súbito resfriamento da parede do vaso. As tensões térmicas, resultantes deste choque térmico, associadas às tensões causadas pela repressurização do sistema, resultam em tensões de tração bastante elevadas, atingindo um valor máximo na superfície interna da parede do vaso. Além disso, a baixa temperatura provoca uma redução na tenacidade à fratura do material. Tal cenário pode levar à propagação de trincas relativamente pequenas através da parede do vaso. Portanto, ferramentas para prever o comportamento de trincas durante um evento de PTS são importantes e necessárias. O tema do presente trabalho se insere neste contexto. Em primeiro lugar, foi feito um estudo das principais questões envolvidas com o problema de PTS em vasos de pressão de reatores PWR. Essas questões dizem respeito ao comportamento à fratura de aços ferríticos na região de transição frágil-dúctil, aos procedimentos de análise de PTS disponíveis em documentos normativos e ao uso de ferramentas de análise numérica para cálculo de distribuição de temperaturas e tensões, e para obtenção de parâmetro de mecânica da fratura representativo da força motriz da trinca. Como principal objetivo do trabalho, foram desenvolvidos modelos de elementos finitos para avaliação do comportamento estrutural de um protótipo de VPR, contendo trincas em sua superfície, utilizado em um experimento de PTS. Procedimentos de mecânica da fratura foram também aplicados para prever eventuais crescimentos de trinca através da espessura da parede do vaso. Resultados das análises numéricas foram comparados com aqueles obtidos com o uso de método simplificado e com medições realizadas no experimento de PTS. / In the primary system of a pressurized water reactor (PWR) nuclear power plant, the reactor coolant is kept at internal temperature around 300 C and internal pressure in the order of 15,0 MPa, during normal operation. The reactor pressure vessel (RPV) contains the fuel assemblies and is considered the most important component of the reactor primary system. The RPV integrity must be assured all along its useful life to protect the general public against radiation liberation damage. One of the most severe load conditions that may threaten the integrity of a RPV is caused by a transient known as pressurized thermal shock (PTS). The RPV may be subjected to such a condition during a loss of coolant accident. In an event like that, the emergency core cooling system is activated, what leads to a sudden cooling of the RPV wall. The thermal stresses due to this thermal shock on the vessel wall, in combination with the pressure stresses from repressurization of the system, results in large tensile stresses, which are maximum at the inside surface of the vessel. In addition, the low temperature causes a decrease in the material fracture toughness. Such a scenario may lead to the propagation of relatively small cracks through the vessel wall. Therefore, analysis tools to predict crack growth behavior during a PTS event are important and necessary. The theme of the present work is connected with this research area. In the first place, the critical issues involved with the PTS problem were reviewed. These issues are related to the fracture behavior of ferritic steels in the ductile-to-brittle transition region, the PTS analysis procedures available in industry codes and standards, and the use of numerical analysis tools for calculation of temperature and stress distribution and for computation of crack driving force parameter. As the main goal, finite element models were developed for the assessment of the structural behavior of a RPV prototype, containing surface cracks, used in a PTS experiment. Fracture mechanics procedures were applied to predict crack growth through the vessel wall. The results of numerical analyses were compared with those obtained with the use of a simplified methodology and measurements from the PTS experiment.