The thesis deals with the perspective of the Austrian newspapers like Die Presse and Kurier on the Czech-Austrian relationship between years 1999 and 2004, or more precisely the issue of Temelín NPP. The main topics for the Austrian press became the approval of completion by the Czech government and of course the launch of Temelín NPP. Significant moments were the blockades and demonstrations of Austrian citizens at the Czech-Austrian border crossings, Austria's attempt to veto the accession negotiations between Czech Republic and EU or internal political struggle in the Austrian government.

The safety of nuclear power plants is still an important topic. Protection of nuclear power plants in Czech Republic is on top level, but there is still place for improving. After the Fukushima disaster Czech Republic responded among other things with creating of new security groups on its own nuclear power plants Temelín and Dukovany. It is rescue group Delta team, specialized in heights security of buildings and Search and Rescue in case of some emergency events. Objective of this Master thesis was preparation and evaluation of practice, where specialized activities of Delta team can be tested. In this case it was ability to give first aid and safe transport of a person from height to the ground using climbing techniques. Communication was added to these tested activities because it is important in every event. The practice was planned and after obtaining of needed permission it was realized on April 1st, 2018 on the nuclear power plant Temelín. The practice was evaluated afterwards, what made objective of this Master theses complete. There is research question in this work and it is if rescue group Delta team is prepared for emergency events related to rescue works in heights. It can be answered positively. Especially the part of practice with climbing operations was precise only with little complaints. This Master thesis serves as familiarization material with the rescue group Delta team, which operates on Czech nuclear power plants. It can give basic information for example for job applicants or for persons interested in nuclear power plant security. Prepared practice can be used repeatedly with other participants or for practice of Delta team on nuclear power plant Dukovany.

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No. of bitstreams: 1 quintella_lc_me_bauru.pdf: 4453734 bytes, checksum: b3c4369bb61bf3c735292c7280dfc961 (MD5) Previous issue date: 2016-07-04 / A função manutenção vem sendo considerada como fator estratégico para as empresas, pois através do alinhamento de suas políticas corporativas e integração de seus programas de gestão de ativos, de riscos e de ciclo de vida de suas unidades de negócios, as empresas vêm buscando a constante redução de custos e a melhoria de seus resultados operacionais. E, assim, obtendo maior competitividade. A Manutenção Centrada em Confiabilidade (RCM) é um método já bem disseminado por todo o mundo e que, ao longo dos anos, vem promovendo estes diferenciais estratégicos através de preceitos que possibilitam a elaboração de Programas de Manutenção Preventiva de custo-eficaz, através de um método para a definição de políticas de manutenção mais adequadas, com o foco na manutenção da função dos ativos em seu contexto operacional. Ao longo dos anos, o método RCM vem sendo aplicado em inúmeros estudos de casos em diferentes empresas de diversos seguimentos, onde podem ser observadas novas adaptações ou simplificações do método RCM clássico. Estas adaptações buscam uma maior adequação as particularidades destas empresas e/ou um retorno mais rápido de resultados. O setor nuclear de geração de energia foi um dos pioneiros na adoção e disseminação do RCM, e vem desenvolvendo processos simplificados de aplicação do RCM, como o “Streamlinned RCM” e o “método PMO” (Otimização do Programa de Manutenção, do inglês: Preventive Maintenance Optimization). Estes estudos mostram que o método PMO apresenta uma maior flexibilidade, o que permite a adoção de diferentes estratégias de aplicação que, por sua vez, têm trazido resultados expressivos para as empresas, através da otimização dos Programas de Manutenção já existentes. Com base na literatura, neste trabalho são abordadas questões referentes ao RCM e sua contextualização na área nuclear, estudos sobre os métodos simplificados do RCM e o desenvolvimento do método PMO. Por fim, é realizada uma aplicação prática do método PMO sobre os sistemas relacionados e subsistemas do Sistema de Remoção de Calor Residual (JN), mais especificamente, sobre as Bombas de injeção de segurança do Sistema de Injeção de Alta Pressão (JND) da Usina Nuclear Angra 2. Através dos resultados obtidos com esta aplicação, pretende-se otimizar o Programa de Manutenção da Planta (PMP) referente a estes equipamentos e, assim, validar o método PMO como ferramenta para a melhoria contínua do Programa de Gestão de Ativos e Ciclo de Vida da Usina Nuclear Angra 2. / Corporations tend to consider maintenance work a strategic element. It is through maintenance—especially the alignment between corporate policies and integration of their programs to manage assets, risks, and life cycles of business units—that corporations try to continuously improve their results, reduce their operational costs, and, therefore, increase their competitiveness. Reliability Centered Maintenance (RCM) is a popular method across the world; it has been promoting competitiveness through concepts that allow the design of cost-effective Preventive Maintenance Programs These programs are effective because they entail appropriate maintenance policies that are focused on the preservation of the assets functions in their operational context and on the formation of technical knowledge basis supported by hard data. Throughout the years, RCM has been applied in numerous case studies and in different companies engaged in a variety of market segments. Such diversity in the application of RCM allowed us to observe new adaptations and variations of the classic method. Such adaptations aim to better respond to specific operational contexts in and within production units, as well as achieve faster results. The nuclear power sector has pioneered regarding the adoption and dissemination of RCM; it has been developing simplified versions of RCM, such as the “Streamlined RCM” and the PMO (Preventive Maintenance Optimization). These studies demonstrate that the PMO presents enhanced flexibility, which allows the adoption of different strategies; such enhanced flexibility brings expressive results to corporations as pre-existing maintenance programs are optimized. Based on the currently available literature, this dissertation addresses numerous questions regarding RCM and its application to nuclear power segments. It also addresses studies about simplified versions of RCM and the development of the PMO method. The discussion is supplemented with a practical application of the PMO method regarding auxiliary systems and sub-systems of Removal of Residual Heat System (JN), especially those regarding the security injection pumps of the High Pressure Injection System (JND) at Angra II Nuclear Plant. Through the results obtained from this application, it is possible to optimize the Maintenance Program of these equipments, and therefore, validate the PMO method as a tool of continuous improvement of the Assets and Life Cycle Program of Angra II.

Nuclear facility accidents involving leaks of radioactive substances into the environment might have serious radiobiological impacts on inhabitants in some cases and would require immediate or subsequent protective measures from iodine prophylaxis and hiding to evacuation or permanent resettlement. The first part of the thesis describes the conditions of nuclear facility operation, technology of the Temelin nuclear power plant, including description of technology and function of emergency systems and the basic principles of security, and environment protection. It also describes the principles and aims of nuclear safety, radiation protection and accident preparedness. A more detailed description of accident preparedness at the Temelin power plant including links to the External Emergency Plan and announcement of safety measures for inhabitants follows. The aim of the thesis is to evaluate the decision making process on announcement of evacuation in the pre-leak phase of radiation accident from the point of view of the Temelin nuclear plant operator?s possibilities. The part Methodology includes a description of RTARC software for evaluation of radiological impacts of radioactive substance leakage into the environment of the Temelin nuclear plant in case of radiation accident. The set of source members (event scenarios) used for the calculation is also described there. The chapter Results presents and assesses the RTARC calculations. We may say upon analyses of the results that application of the decision making support programme is beneficial and is practically able to help with efficient protection of citizens in the final effect. For a part of the considered source members evacuation on pre-leak phase cannot be recommended with regard to the quick progress and leakage of radioactive substances into the environment.

Advanced computational tools are applied to simulate a nuclear power plant (NPP) control rod assembly ejection (CRE) accident. The impact of these reactivity-initiated accidents (RIAs) on core reactivity behavior, 3D power distribution and stochastic reactivity estimation are evaluated. The three tools used are: the thermal-hydraulic (TH) RELAP5 (R5) code, the neutronic (NK) PARCS (P3D) code, and the coupled version P3D/R5, with specially developed linkage using the environment code MATLAB. This study considers three different-size cores: NPP1 (2772 MWt); NPP2 (530 MWt); and NPP3 (1061 MWt). The three cores have the same general design and control rod assembly (CRA) positions, and the ejected CRA has similar worth and at the same rod ejection pace. The CRE is assessed under both hot zero power (HZP) and hot full power (HFP) conditions. The analyses indicate that RIA modeling and simulation should be carried out through a systematic coding and configuration approaches, otherwise the results will not capture the true transient behavior of the core under analysis. The simulation of one code depends on the appropriate configuration of parameters generated by the other code and on the correct determination of the TH/NK mapping weight factors for the various mesh regions in each of the models. From the design point of view, the standalone codes predict milder magnitude of power and reactivity increase compared to the coupled P3D/R5 simulation. The magnitudes of reduced peak power and reactivity become larger as the core size shrinks. The HFP simulation shows that the three NPPs have the same transient peak value, but the post-transient steady power is lower for a smaller core. The HZP analysis indicates that the transient peak is lower for the smaller core, but the post-transient power occurs at the same level. The three-dimensional (3D) power distributions are different among the HFP and HZP cases, but do not depend on the size of the core. The results indicate: i) HFP: core power increases in the area surrounding the ejected rod/bank assembly, and this increase becomes lower as the NPPs shrinks however, the power is well-distributed after the transient; and ii) HZP: the area surrounding the CRA stays hotter, but the 3D peak assembly factor becomes lower, during and after the transients, as the NPPs shrinks. These features confirm that the smaller cores yield a safer response to a given inserted reactivity compared to larger cores. A stochastic extended Kalman filter (EKF) algorithm is implemented to estimate the reactivity based on the reactor power profile, after the addition of random noise. The inverse point kinetics (IPK) deterministic method is also implemented and the results of the application of EKF and IPK are compared to the P3D/R5 simulation. The following sophisticated strategies made the EKF algorithm robust and accurate: the system is modeled by a set of continuous time nonlinear stochastic differential equations; the code uses a time step directly based on the power measured and applies that to the model for online discretization and linearization; filter tuning goes automatically up from the first time step; and the state noise covariance matrix is updated online at each time step. It was found that the IPK reactivity has higher noise content compared to the EKF reactivity for all cases. Thus, the EKF presents superior and more accurate results. Furthermore, under a small reactivity insertion, the IPK reactivity varies widely from positive to negative values: this variation is not observed within the EKF. A sensitivity analysis for three distinct standard deviation (SD) noise measurements suggests that EKF is superior to IPK method, independent of the noise load magnitude. As the noise content increases, the error between the IPK and P3D/R5 reactivity also increases. A sensitivity analysis for five distinct carry-over effects of different random noise loads indicates that the random addition of different noise loads to the reactor power does not change the overall performance of both algorithms. / Este trabalho aplica métodos computacionais avançados para simular a ejeção de barras de controle (CRE) em uma planta térmica nuclear (NPP). São avaliados o impacto da ocorrência de acidentes iniciados por reatividade (RIAs) na reatividade total, na distribuição da potência em três dimensões (3D) e na determinação da reatividade. As ferramentas utilizadas são: o código termo-hidráulico (TH) RELAP5 (R5), o código neutrônico (NK) PARCS (P3D), a versão acoplada P3D/R5, e o ambiente computacional MATLAB. Este estudo considera três reatores nucleares de diferentes tamanhos: NPP1 (2772 MWT); NPP2 (530 MWt); e NPP3 (1061 MWt). Os três núcleos possuem projeto similar e idêntica posição dos grupos das barras de controle (CRA), além do mesmo valor de reatividade diferencial das CRA ejetadas e idêntica velocidade de ejeção. A ocorrência da CRE é avaliada sob condições de hot zero power (HZP) e de hot full power (HFP). As análises indicam que a modelagem e a simulação de RIAs devem ser realizadas sistematicamente, caso contrário os resultados não irão refletir o comportamento em regime transitório do núcleo. A simulação de um modelo em um código depende da apropriada configuração de parâmetros gerados pelo outro código e da determinação adequada do mapeamento TH/NK para as várias malhas dos modelos. Do ponto de vista de projeto, a utilização de códigos independentes resulta em cálculos de potência e reatividade conservadores em comparação com os resultados utilizando-se P3D/R5. Os picos de potência e de reatividade são menores à medida que o núcleo encolhe. A simulação em condições de HFP resulta em valores de pico de potência similares durante transitório para as três NPPs, mas a potência de pós-transitórios é menor para o menor núcleo. A análise em condições de HZP também indica que o valor máximo durante o transitório é menor para o menor núcleo, mas o pós-transitórios ocorre aos mesmos níveis de potência das demais NPPS. A distribuição de potência em 3D também apresenta resultados distintos para condições de HFP e HZP, mas tais resultados são independentes do tamanho do núcleo: i) HFP: há um aumento da potência do núcleo em torno da CRE, mas tal comportamento diminui para núcleos menores - no entanto, a potência é bem distribuída após o transitório; e ii) HZP: há aumento de potência na área do CRE, mas o pico de potência em 3D é menor durante e depois dos transitórios para núcleos menores. Tais características indicam que os núcleos menores respondem de forma mais segura quando da inserção de reatividade em comparação a reatores de maiores dimensões. O método estocástico de filtragem de Kalman estendido (EKF) foi codificado para estimar a reatividade com base no perfil de potência da NPP, após a adição de ruído aleatório. O método determinístico da cinética pontual inversa (IPK) também foi implementado e os resultados da aplicação dos algoritmos do EKF e IPK foram comparados com os resultados da simulação do P3D/R5. As seguintes estratégias, implementadas neste trabalho, possibilitaram a aplicação robusta e precisa do EKF: o sistema foi modelado por um conjunto de equações diferenciais não-lineares estocásticas de tempo contínuo; o algoritmo obtém o passo de tempo diretamente da potência medida e aplica-o ao modelo para a discretização e linearização online; o ajuste do filtro ocorre automaticamente a partir do primeiro passo de tempo; e a matriz de covariância do ruído no estado é atualizada online. Verificou-se que a reatividade calculada pelo método IPK possui maior nível de ruído quando comparada ao EKF para todos os casos estudados. Portanto, o EKF apresenta resultados superiores e mais precisos. Além disso, sob uma pequena inserção de reatividade, a reatividade calculada pelo método IPK varia consideravelmente de valores positivos para negativos: esta variação não é observada com o EKF. Uma análise de sensibilidade para três desvios padrão (SD) sugere que o algoritmo EKF é superior ao método IPK, independente da magnitude do ruído. Com o aumento da magnitude do ruído, o erro entre as reatividades calculadas pelo IPK e pelo P3D/R5 aumenta. A análise de sensibilidade para cinco ruídos aleatórios indica que a adição de ruído na potência do reator não altera o desempenho global de ambos os algoritmos.

Nuclear power plays a crucial role in energy supply in the world: around 15% of the electricity generated worldwide is provided from nuclear stations avoiding around 2.5 billion tonnes of CO2 emissions. As of January 2016, 442 reactors that generated 380+ GW were in operation and 66 new reactors were under construction. The seismic design of new nuclear power plants (NPPs) has gained much interest after the high-profile Fukushima Dai-ichi accident. In the UK, a tectonically stable continental region that possesses medium-to-low seismic activity, strong earthquakes capable of jeopardising the structural integrity of NPPs, although infrequent, can still occur. Despite that no NPP has been built in Great Britain after 1995, a New Build Programme intended to build 16 GW of new nuclear capacity by 2030 is currently under way. This PhD project provides a state-of-the-art framework for seismic probabilistic safety assessment and risk control of NPPs in Northwest Europe with particular application to the British Isles. It includes three progressive levels: (i) seismic input, (ii) seismic risk analysis, and (iii) seismic risk control. For seismic input, a suitable model to rationally define inputs in the context of risk assessments is proposed. Such a model is based on the stochastic simulation of accelerograms that are compatible with seismic scenarios defined by magnitude 4 < Mw < 6.5, epicentral distance 10 km < Repi < 100 km, and different types of soil (rock, stiff soil and soft soil). It was found to be a rational approach that streamlines the simulation of accelerograms to conduct nonlinear dynamic analyses for safety assessments. The model is a function of a few variables customarily known in structural engineering projects. In terms of PGA, PGV and spectral accelerations, the simulated accelerograms were validated by GMPEs calibrated for the UK, Europe and the Middle East, and other stable continental regions. For seismic risk analysis, a straightforward and logical approach to probabilistically assess the risk of NPPs based on the stochastic simulation of accelerograms is studied. It effectively simplifies traditional approaches: for seismic inputs, it avoids the use of selecting/scaling procedures and GMPEs; for structural outputs, it does not use Monte Carlo algorithms to simulate the damage state. However, it demands more expensive computational resources as a large number of nonlinear dynamic analyses are needed. For seismic risk control, strategies to control the risk using seismic protection systems are analysed. This is based on recent experience reported elsewhere of seismically protected nuclear reactor buildings in other areas of medium-to-low seismic activity. Finally, a scenario-based incremental dynamic analysis (IDA) is proposed aimed at the generation of surfaces for unacceptable performance of NPPs as function of earthquake magnitude and distance. It was found that viscous-based devices are more efficient than hysteretic-based devices in controlling the seismic risk of NPPs in the UK. Finally, using the proposed scenario-based IDA, it was found that when considering all controlling scenarios for a representative UK nuclear site, the risk is significantly reduced ranging from 3 to 5 orders of magnitude when using viscous-based devices.

Rapporten baseras på en litteratur- och intervjustudie kring underhållsstrategier och komponentutbyten i kärnkraftverk, med fokus på komponenter som tenderar till att påverkas av åldring i en högre grad. Exemplifiering sker genom en kompletterande fallstudie kring kabel- och rörgenomföringar av typen Brattbergare som har packbitar bestående av ett polymert material. Erhållet resultat av litteratur- och intervjustudien belyser vilka säkerhetsföreskrifter som råder för all kärnteknisk verksamhet i Sverige. Utöver det erhålls information om hur åldring påverkar ett materials egenskaper över tid och att detta ligger till grund för fastställandet av ett system eller en komponents kvalificerade livslängd. I takt med att majoriteten av världens kärnkraftverk närmar sig sin ursprungligt tilltänkta livslängd och planeras underhållas för fortsatt långtidsdrift finns det ett ökat intresse för effektiva underhållsstrategier. Åldershanteringen har en avgörande roll för anläggningens lönsamhet och driftsäkerhet. Fallstudien föreslår två olika underhållsstrategier som stöds av resultatet av litteratur- och intervjustudien. Deras ekologiska påverkan och ekonomiska omfattning beaktas för att utse den metod som har störst potential att öka resurseffektiviteten och minska kostnaderna för underhållsåtgärder. Vald metod går ut på att praktiska tillståndsmätningar tillämpas för att undersöka hårdheten av packbitar till Brattbergare. Hårdhetsmätningarna syftar till att ge indikationer på i vilken grad packbitarna harpåverkats av degraderande åldringsmekanismer under olika förutsättningar. Resultatet av fallstudien överensstämmer med de resultat som noterades i litteratur- och intervjustudien. Packbitarna hårdnar när de åldras. Två miljöbetingelser som tenderar till att påskynda åldringsprocessen är förhöjda temperaturer och stråldoser. Vald metod för fallstudien är praktiskt realiserbar trots vissa begränsningar i befintliga kärnkraftsanläggningar vid Forsmark. Presenterad strategi bör kunna bistå med en ekologisk och ekonomisk optimering av underhållsarbetet för kabel- och rörgenomföringar. / This report is based on a literature study and interviews regarding maintenance strategies and component replacements in nuclear power plants. Focuses of the study are on components which tend to more commonly be affected by degrading aging mechanisms. Exemplification is done with a complementary case study about cable- and pipe transits with packing pieces made of polymeric materials. A frequently used application for cable- and pipe transits in Swedish NPPs is manufactured by MCT Brattberg AB.  Result obtained from interviews with relevant personnel’s and the literature study providing knowledge about prevailing safety regulations at Swedish nuclear facilities. Moreover, information is gained about how aging affects the features of materials over time and that it is the basis for determining the qualified lifetime of systems and components. Further on this could be of specific interest considering that the majority of the worlds NPPs are close to its intended lifetime and soon entering a phase of LTO. A proper aging management is an important factor when it comes to a safe and reliable operation of an NPP.  The case study compares two different maintenance strategies which are supported by the obtained result from interviews and the literature study. Considering ecological and economic impacts of the strategies, the one with the greatest potential to reduce negative influences are exemplified. Chosen method included practical hardness measurement with a portable durometer at packing pieces for cable- and pipe transits. Measured hardness of the packing pieces indicates at what degree which they have been affected by degrading aging mechanisms given different circumstances.   The obtained result from the two different parts of the report is corresponding to each other. Packing pieces consisting polymers hardens as they age. Elevated temperatures and higher dose rates accelerates the aging process. Represented method of the case study is practically viable at existing NPPs at Forsmark. Presented strategy should be able to assist with an ecological and economic optimization maintenance work for cable- and pipe transits.

The thesis is focused on multiparametric diagnostic of generators at the Dukovany nuclear power plant. One generator was chosen for the thesis and it was examined especially from the practical point of view. The thesis describes current state of the issue with focus on noise diagnostics, vibrodiagnostics and electrodiagnostics. The emphasis is on the system approach of the solution. In the practical part an experiment is designed, described and evaluated. Attention is paid to the conclusions of the performed vibrodiagnostics and noise diagnostics. Data sets are evaluated separately and the relationship between the two diagnostic methods is analyzed. Furthermore, the data set from electrodiagnostics is evaluated and dependencies of electrical diagnostic quantities on other quantities are described. Trends in electrical diagnostic quantities are also monitored. Conclusions and recommendations are formulated at the end of the thesis. It is stated that using multiparametric diagnostics to assess the status of generators in power plants is still in its beginning and that the conclusions of the submitted thesis will contribute to the developmnet in this area.

The subject of the diploma thesis is the elaboration of an architectural study of the reconstruction of the pre-plant zone of the Facelift of the Dukovany power plant. The construction site is an area that belongs to the village of Dukovany and borders the village of Rouchovany. As a whole, this area is in poor technical and architectural condition. However, its location gives potential for new uses. The solved area belongs to the ČEZ nuclear power plant.

The subject of the diploma thesis is the elaboration of an architectural study of the reconstruction of the pre-plant zone of the Facelift of the Dukovany power plant. The construction site is an area that belongs to the village of Dukovany and borders the village of Rouchovany. As a whole, this area is in poor technical and architectural condition. However, its location gives potential for new uses. The solved area belongs to the ČEZ nuclear power plant.