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.

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.

A importância de métodos de projeto vem crescendo nos últimos anos à medida que sistemas sócio-técnicos complexos se tornam mais numerosos. Além da complexidade, o tamanho e o investimento financeiro destes sistemas amplificam a gravidade dos erros de projeto. O objetivo geral deste trabalho foi desenvolver uma metodologia de especificação e projeto que reduza o tempo e energia para desenvolver um sistema complexo cujas funções sejam conhecidas a priori, gerenciando em paralelo os riscos. O objetivo específico foi verificar a viabilidade econômica de usinas nucleares móveis de pequeno porte. Este trabalho adotou como princípio a lei construtal que prevê o sucesso de sistemas que facilitem os fluxos necessários à sua existência. Após a identificação dos fatores chave para facilitar o fluxo de informações, esta tese desenvolveu um conjunto de conceitos para facilitar o trabalho de engenharia. Aplicando tais conceitos, este trabalho desenvolveu sequências de atividades que descrevem o método proposto, sendo cada atividade detalhada por uma lista de requisitos. A demonstração das vantagens do método proposto foi feita por meio de análise de árvore de eventos e árvore de falhas. Usando o método, esta tese desenvolveu especificações e projetos em vários níveis (empresarial, usina, caldeira nuclear, circuito primário e gerador de vapor). Baseando-se em dados da marinha americana, esta tese desenvolveu um modelo de custo para reatores de pequeno porte. Concluiu que a energia nuclear pode ser competitiva se a potência elétrica média efetiva ao longo da vida útil ficar acima de 30MWe e se o tempo de vida útil for igual ou maior do que 60 anos. Tal fato decorre dos altos custos de aquisição que requerem uma vida longa para compensar o investimento e dos efeitos de economia de escala especialmente pronunciados para reatores a água pressurizada. / The importance of design methodologies has been growing in recent years as complex socio-technical systems become more common. In addition to complexity, the size and financial investment of these systems amplifies the severity of design errors. The general goal of this work was to develop a specification and design methodology that reduces the time and energy to develop a complex system whose functions are known a priori, managing the risks in parallel. The specific goal was to verify if small modular reactors could be economically possible. This work adopted as principle the Constructal law, that predicts the success of systems that ease the necessary flows to its existence. After finding the key factors to ease the flow of information, this work developed a set of concepts to ease the engineering work. Applying such concepts, this work developed sequences of activities that describe the proposed methodology. Lists of requirements gave guidance for each activity. Event tree and fault tree analyses showed the advantages of the proposed methodology. Using the methodology, this work developed specifications and designs at many product breakdown levels (enterprise, nuclear power plant, nuclear steam supply system, reactor coolant system and steam generator). Using data from US Navy, this work developed a cost model for small reactors. This work concluded that nuclear power may be competitive if average electrical power extracted during the life is larger than 30 MWe and if life time is superior to 60 years. The first condition is consequence of the high overnight costs of nuclear power. The second is consequence of the strong scale economy effects of pressurized water reactors.

<p>Kärnkraften ser vi som ett miljövänligt och bra alternativ för framtiden. Därför kändes det intressant att undersöka om vi som blivande sjöingenjörer har en framtid där efter våran karriär till sjöss. Syftet med detta arbete är att få ökad insikt om vilka teoretiska kunskaper, praktiska och personliga egenskaper som gör sjöingenjören attraktiv i ledande befattningar vid kärnkraftsverk. Datainsamlingen skedde med hjälp av en enkätundersökning och en intervju. Undersökningen skickades ut till Sveriges samtliga kärnkraftverk och intervjun gjordes med en anställd på ett utav dem. Detta citat ifrån en av undersökningens respondenter kan sammanfatta vår slutsats. ” Sjöingenjörsprogrammet är en mycket attraktiv utbildning för oss!”. Detta resultat, kan bero på den stora bristen och ointresse från unga att läsa till ingenjörsyrken, samt ingenjörernas höga medelålder och kommande pensionsavgångar.</p> / <p>It is our belief that nuclear power constitutes a future environmentally friendly alternative power source. Based on this conviction, our study was conducted with the principal aim of investigating whether or not a marine engineer would be eligible for employment at a nuclear power plant in Sweden after having ended his sea-going career.</p><p>The primary objective of our investigation was to examine what type of theoretical knowledge that the nuclear industry required. Furthermore, we also wanted to find answers to questions related to such requirements as practical and personal characteristics. Overall, we wanted to find out what personal and educational qualities that were needed to make marine engineers attractive as prospective employees for the nuclear power industry.</p><p>The collection of data was primarily made with the help of a questionnaire, which was sent to all nuclear power plants in Sweden. We also conducted an in-depth interview with one nuclear power plant employee.</p><p>The conclusion of our investigation into this field of possible future employment for a marine engineer can be deduced from the answer given by one of the questionnaire respondents: “The marine engineering programme is a very attractive education for us.”</p><p>The latter answer together with our general conclusions from the questionnaire, substantiate the fact that marine engineers will be plausible candidates for employment at nuclear power plants in the future.</p>

Kärnkraften ser vi som ett miljövänligt och bra alternativ för framtiden. Därför kändes det intressant att undersöka om vi som blivande sjöingenjörer har en framtid där efter våran karriär till sjöss. Syftet med detta arbete är att få ökad insikt om vilka teoretiska kunskaper, praktiska och personliga egenskaper som gör sjöingenjören attraktiv i ledande befattningar vid kärnkraftsverk. Datainsamlingen skedde med hjälp av en enkätundersökning och en intervju. Undersökningen skickades ut till Sveriges samtliga kärnkraftverk och intervjun gjordes med en anställd på ett utav dem. Detta citat ifrån en av undersökningens respondenter kan sammanfatta vår slutsats. ” Sjöingenjörsprogrammet är en mycket attraktiv utbildning för oss!”. Detta resultat, kan bero på den stora bristen och ointresse från unga att läsa till ingenjörsyrken, samt ingenjörernas höga medelålder och kommande pensionsavgångar. / It is our belief that nuclear power constitutes a future environmentally friendly alternative power source. Based on this conviction, our study was conducted with the principal aim of investigating whether or not a marine engineer would be eligible for employment at a nuclear power plant in Sweden after having ended his sea-going career. The primary objective of our investigation was to examine what type of theoretical knowledge that the nuclear industry required. Furthermore, we also wanted to find answers to questions related to such requirements as practical and personal characteristics. Overall, we wanted to find out what personal and educational qualities that were needed to make marine engineers attractive as prospective employees for the nuclear power industry. The collection of data was primarily made with the help of a questionnaire, which was sent to all nuclear power plants in Sweden. We also conducted an in-depth interview with one nuclear power plant employee. The conclusion of our investigation into this field of possible future employment for a marine engineer can be deduced from the answer given by one of the questionnaire respondents: “The marine engineering programme is a very attractive education for us.” The latter answer together with our general conclusions from the questionnaire, substantiate the fact that marine engineers will be plausible candidates for employment at nuclear power plants in the future.

In the event of a nuclear power plant accident, fission products could be released into the atmosphere potentially affecting the health of local citizens. In order to triage the possibly large number of people impacted, a detection device is needed that can acquire data quickly and that is sensitive to internal contamination. The portal monitor TPM-903B was investigated for use in the event of a fission product release. A list of fission products released from a Pressurized Water Reactor (PWR) was generated and separated into two groups--Group 1 (gamma- and beta-emitting fission products) and Group 2 (strictly beta-emitting fission products.) Group one fission products were used in the previously validated Monte Carlo N-Particle Transport Code (MCNP) model of the portal monitor. Two MIRD anthropomorphic phantom types were implemented in the MCNP model--the Adipose Male and Child phantoms. Dose and Risk Calculation software (DCAL) provided inhalation biokinetic data that were applied to the output of the MCNP modeling to determine the radionuclide concentrations in each organ as a function of time. For each phantom type, these data were used to determine the total body counts associated with each individual gamma-emitting fission product. Corresponding adult and child dose coefficients were implemented to determine the total body counts per 250 mSv. A weighted sum of all of the isotopes involved was performed. The ratio of dose associated with gamma-emitting fission products to the total of all fission products was determined based on corresponding dose coefficients and relative abundance. This ratio was used to project the total body counts corresponding to 250mSv for the entire fission product release inhalation--including all types of radiation. The developed procedure sheets will be used by first response personnel in the event of a fission product release.

Thesis (M. S.)--Mechanical Engineering, Georgia Institute of Technology, 2008. / Committee Chair: Willem F.G. Van Rooijen; Committee Member: Ghiaasiaan, Seyed M; Committee Member: Weston M. Stacey

[pt] Os resultados apresentados referem-se à investigação dos mecanismos de interação entre dois sistemas vitais às usinas nucleares, ou seja: os sistemas Principal (SP) e o Secundário (SS). Estes mecanismos são avaliados por meio de sua influência nos espectros de resposta, em pontos da estrutura passíveis da existência de suportes das linhas de tubulação - SS. São usados dois tipos diferentes de análises para a geração dos espectros de resposta: a primeira não considera a interação dos sistemas e a segunda avalia esta interação com a introdução, em cada ponto de suporte no SP, de um S1GL com suportes únicos ou com multi- suportes. As respostas estruturais são obtidas por integração direta da equação de movimento do sistema sujeito a dois acelerogramas simultâneos, nas direções horizontal e vertical. Os resultados são analisados e comparados para identificação das principais tendências das análises e esclarecimento dos efeitos envolvidos. Estuda-se, também, a importância de não- linearidades concentradas nos suportes da tubulação, tendo- se em vista o nível sísmico a que as centrais nucleares brasileiras estão sujeitas. / [en] The results presented in this work refer to the investigation of the mechanics of the interaction between two important systems of nuclear power plants, i.e.: the Primary (PS) and Secondary (SS) systems. The influence of these effects on the response spectra is studied, in convenient points of the structure where could exist pipeline (SS) supports. Two different approaches are used to generate the response spectra: the first neglects the interaction between the two systems and the second considers this interaction by the addition, to every support point on the PS, of a single-supported or multi-supported SDOF system. The structural responses are obtained by the direct integration of the Primary System equations of motion subjected to two simultaneous design acceleration time-histories, in the horizontal and vertical directions. The results are analyzed and compared to identify the general trends of the solutions obtained by the two types of analysis, and to detect their effects on the SS response. The study is concerned, also, with the importance of nonlinearities concentrated in the pipeline supports; in the case of the Brazilian nuclear power plants. / [es] Los resultados presentados se refieran a la investigación de los mecanismos de interacción entre de los los sistemas Principal (SP) y el Secundario (S) de las plantas nucleares. Estos mecanismos son evaluados por medio de su influencia en los espectros de respuesta, en puntos de la extructura donde es posible(pausibles) la existencia de soportes de las líneas de tuberías - S. Son usados dos tipos diferentes de análisis para la generación de los espectros de respuesta: la primera no considera la interacción de los sistemas y la segunda evalúa esta interacción con la introdución, en cada ponto de soporte en el SP, de un S1GL con soportes únicos o con multisoportes. Las respuestas extructurales son obtenidas por integración directa de la ecuación de movimento del sistema sujeto a dos acelerogramas simultáneos, en las direcciones horizontal y vertical. Se analizan los resultados y se comparan para identificar las principales tendencias del análisis y esclarecer los efectos involucrados. Se estudia además, la importancia de no linealidades concentradas en los soportes de la tubería, teniendo en vista el nível sísmico a que las centrales nucleares brasileras están sujetas.

A importância de métodos de projeto vem crescendo nos últimos anos à medida que sistemas sócio-técnicos complexos se tornam mais numerosos. Além da complexidade, o tamanho e o investimento financeiro destes sistemas amplificam a gravidade dos erros de projeto. O objetivo geral deste trabalho foi desenvolver uma metodologia de especificação e projeto que reduza o tempo e energia para desenvolver um sistema complexo cujas funções sejam conhecidas a priori, gerenciando em paralelo os riscos. O objetivo específico foi verificar a viabilidade econômica de usinas nucleares móveis de pequeno porte. Este trabalho adotou como princípio a lei construtal que prevê o sucesso de sistemas que facilitem os fluxos necessários à sua existência. Após a identificação dos fatores chave para facilitar o fluxo de informações, esta tese desenvolveu um conjunto de conceitos para facilitar o trabalho de engenharia. Aplicando tais conceitos, este trabalho desenvolveu sequências de atividades que descrevem o método proposto, sendo cada atividade detalhada por uma lista de requisitos. A demonstração das vantagens do método proposto foi feita por meio de análise de árvore de eventos e árvore de falhas. Usando o método, esta tese desenvolveu especificações e projetos em vários níveis (empresarial, usina, caldeira nuclear, circuito primário e gerador de vapor). Baseando-se em dados da marinha americana, esta tese desenvolveu um modelo de custo para reatores de pequeno porte. Concluiu que a energia nuclear pode ser competitiva se a potência elétrica média efetiva ao longo da vida útil ficar acima de 30MWe e se o tempo de vida útil for igual ou maior do que 60 anos. Tal fato decorre dos altos custos de aquisição que requerem uma vida longa para compensar o investimento e dos efeitos de economia de escala especialmente pronunciados para reatores a água pressurizada. / The importance of design methodologies has been growing in recent years as complex socio-technical systems become more common. In addition to complexity, the size and financial investment of these systems amplifies the severity of design errors. The general goal of this work was to develop a specification and design methodology that reduces the time and energy to develop a complex system whose functions are known a priori, managing the risks in parallel. The specific goal was to verify if small modular reactors could be economically possible. This work adopted as principle the Constructal law, that predicts the success of systems that ease the necessary flows to its existence. After finding the key factors to ease the flow of information, this work developed a set of concepts to ease the engineering work. Applying such concepts, this work developed sequences of activities that describe the proposed methodology. Lists of requirements gave guidance for each activity. Event tree and fault tree analyses showed the advantages of the proposed methodology. Using the methodology, this work developed specifications and designs at many product breakdown levels (enterprise, nuclear power plant, nuclear steam supply system, reactor coolant system and steam generator). Using data from US Navy, this work developed a cost model for small reactors. This work concluded that nuclear power may be competitive if average electrical power extracted during the life is larger than 30 MWe and if life time is superior to 60 years. The first condition is consequence of the high overnight costs of nuclear power. The second is consequence of the strong scale economy effects of pressurized water reactors.

Orientador: Prof. Dr. Pedro Carajilescov / Dissertação (mestrado) - Universidade Federal do ABC. Programa de Pós-Graduação em Energia, 2016. / Os estudos sobre transitórios em bombas de refrigeração de um reator são importantes para a análise de segurança de uma central nuclear. Uma análise precisa do decaimento da vazão de refrigerante no circuito primário durante uma eventual falha das bombas principais de refrigeração, evento conhecido como coastdown, é requerida tanto para pelos critérios de segurança estabelecidos como para a especificação e fabricação das bombas. Neste trabalho, o estudo é realizado utilizando um modelo matemático para simular transiente de vazão durante o período de coastdown em reatores nuclear do tipo PWR, no qual a equação de conservação da quantidade de movimento linear é utilizada de uma forma adimensional. Informações detalhadas sobre as características da bomba centrífuga não são necessárias. Como resultado, o decaimento da vazão é determinado a partir da razão entre dois parâmetros: a energia cinética do fluido refrigerante no circuito e a energia cinética armazenada nas partes rotativas da bomba. Em estudos anteriores, essa razão, denominada razão de energia efetiva, é mantida constante durante todo o evento de coastdown. Neste trabalho, foram propostas três correções para a melhoria dos resultados, a saber: a consideração de uma razão de energia efetiva variável durante o transitório, das variações na eficiência da bomba durante o transitório e das perdas mecânicas internas devido ao atrito e viscosidade no interior da bomba para baixas rotações. Para implementação do modelo proposto foi desenvolvido um programa, denominado de COREP-flow, cujos resultados foram comparados com dados experimentais obtidos na literatura. As comparações mostraram uma melhoria na reprodução desses resultados em relação aos modelos de referência. No Modelo 5 desenvolvido neste trabalho, os resultados obtidos apresentaram menor discrepância quando comparados com os dados experimentais. Para vazões superiores a 20 % da vazão inicial, a vazão de refrigerante calculada pelo Modelo 5 apresentou uma discrepância relativa média de 1,8 %, enquanto que o modelo proposto por Gao et al. (2011) apresentou uma discrepância relativa média de 4 %. Para vazões de refrigerantes inferiores a 20 % da vazão inicial, a discrepância relativa média para a vazão de refrigerante do Modelo 5 foi de 10,3 %, enquanto que a de Gao et al. (2011) foi de 50,6 %. / The transient studies in reactor cooling pumps (RCPs) are important for the nuclear power plant security analysis. An accurate analysis of flow coastdown in the primary cooling loop system during an eventual failure of the RCPs is required both for the established security criteria, and for the pumps specification and manufacturing. In this work, the study is performed using a mathematical model to simulate the flow rate transient in PWR reactor type during flow coastdown period, in which the conservation of linear momentum equation is non-dimensional. The detailed information of the centrifugal pump characteristics are not required. As result, the coastdown is determined from the ratio between two parameters: the kinetic energy of the coolant in the circuit and the kinetic energy stored in the rotating parts of the pump. In previous studies, the ratio, known as energy ratio, is kept constant during the whole coastdown period. In this work, it was proposed three corrections aiming the improvement of the results, to know: the consideration of an energy ratio variable during the transient, of the efficiency variations of the pumps during the transient and of the internal mechanical losses due to friction and viscosity inside the pump for low rotations. For the model implementation, a program was developed, the COREP-flow, whose results were compared to the experimental data obtained in the literature. The comparison showed an improvement in the reproduction of the results in relation to the reference models. In Model 5, developed in this work, the obtained results presented less discrepancy when compared to the experimental data. For flow rates higher than 20% of the initial flow, the coolant flow calculated by Model 5 presented a mean relative discrepancy of 1.8%, while the model proposed by Gao et al. (2011) presented a mean relative discrepancy of 4%. For coolant flow rates less than 20% of the initial flow, the mean relative discrepancy for the coolant flow of Model 5 was of 10.3%, while the one from Gao et al. (2011) presented a mean relative discrepancy of 50.6%.