In order to cope with the changing climate, there will be a need for mitigating transformations of a scope, speed and magnitude that are unprecedented in human history, but the consensus- and market-driven approach is inhibiting this transformation. This thesis reviews how various discourses and debates on energy policy within Swedish governments have changed between 1974 and 2017, through the lenses of energy security and depoliticisation, and how a better understanding of these debates and discourses may inform the impending large-scale transformation required to meet the challenge of climate change. Some of the main findings are that (1) nuclear power and the result of the nuclear power referendum have been decisive for energy policy, and that nuclear power will remain of vital importance in the twenty-first century. (2) Energy has become increasingly depoliticised since the 1980s, ensuing the referendum on nuclear power. (3) The discourse on energy security has shifted towards market-based solutions: in the 2000s, climate change is to be mitigated through consumer- oriented solutions such as green certificates. Through privatisation, it essentially has become up to the consumer, deciding whether to participate in mitigation of climate change.

Submitted by Claudinei Pracidelli (cpracide@ipen.br) on 2016-06-22T14:12:07Z No. of bitstreams: 0 / Made available in DSpace on 2016-06-22T14:12:07Z (GMT). No. of bitstreams: 0 / Após os acidentes nucleares ocorridos no mundo, critérios e requisitos extremamente rígidos para a operação das instalações nucleares foram determinados pelos órgãos internacionais que regulam essas instalações. A partir da ocorrência destes eventos, as operadoras de plantas nucleares necessitam simular alguns acidentes e transientes, por meio de programas computacionais específicos, para obter a licença de operação de uma planta nuclear. Com base neste cenário, algumas ferramentas computacionais sofisticadas têm sido utilizadas como o Reactor Excursion and Leak Analysis Program (RELAP5), que é o código mais utilizado para a análise de acidentes e transientes termo-hidráulicos em reatores nucleares no Brasil e no mundo. Uma das maiores dificuldades na simulação usando o código RELAP5 é a quantidade de informações geométricas da planta necessárias para a análise de acidentes e transientes termo-hidráulicos. Para a preparação de seus dados de entrada é necessário um grande número de operações matemáticas para calcular a geometria dos componentes. Assim, a fim de realizar estes cálculos e preparar dados de entrada para o RELAP5, um pré-processador matemático amigável foi desenvolvido, neste trabalho. O Visual Basic for Applications (VBA), combinado com o Microsoft Excel, foi utilizado e demonstrou ser um instrumento eficiente para executar uma série de tarefas no desenvolvimento desse pré-processador. A fim de atender as necessidades dos usuários do RELAP5, foi desenvolvido o Programa de Cálculo do RELAP5 PCRELAP5 onde foram codificados todos os componentes que constituem o código, neste caso, todos os cartões de entrada inclusive os opcionais de cada um deles foram programados. Adicionalmente, uma versão em inglês foi criada para PCRELAP5. Também um design amigável do PCRELAP5 foi desenvolvido com a finalidade de minimizar o tempo de preparação dos dados de entrada e diminuir os erros cometidos pelos usuários do código RELAP5. Nesse trabalho, a versão final desse pré-processador foi aplicada com sucesso para o Sistema de Injeção de Emergência (SIE) da usina Angra 2. / Dissertação (Mestrado em Tecnologia Nuclear) / IPEN/D / Instituto de Pesquisas Energeticas e Nucleares - IPEN-CNEN/SP

Submitted by Pedro Silva Filho (pfsilva@ipen.br) on 2017-11-21T11:45:58Z No. of bitstreams: 0 / Made available in DSpace on 2017-11-21T11:45:58Z (GMT). No. of bitstreams: 0 / O gerenciamento de riscos em uma instalação nuclear é necessário para a segurança de trabalhadores e de populações vizinhas. Parte desse processo é a comunicação dos riscos que propicia o diálogo entre gestores da empresa e moradores das áreas de risco. A população que conhece os riscos a que está exposta, como esses riscos são gerenciados e o que deve ser feito em uma situação de emergência tende a se sentir mais segura e a confiar nas instituições responsáveis pelo plano de emergência. Sem diálogo entre empresa e público, o conhecimento dos procedimentos a serem seguidos em caso de acidente não chega à população, ou quando chega, não há confiança dessas pessoas na sua eficácia. Em Angra dos Reis, no litoral sul do Estado do Rio de Janeiro, está a Central Nuclear Almirante Álvaro Alberto. No entorno dessa Central Nuclear existe uma população que, de acordo com o Plano de Emergência Externo (PEE/RJ), deverá ser evacuada ou ficar abrigada, caso ocorra um acidente na instalação. Um trabalho de comunicação de riscos entre esses moradores é necessário para que eles conheçam o plano de emergência e os procedimentos corretos para uma situação de emergência, além de buscar esclarecer dúvidas e mitos. Esse trabalho apresenta uma análise da comunicação dos riscos feita para a população local, a percepção que ela tem dos riscos e o grau de conhecimento do plano de emergência externo por parte dessas pessoas. / Tese (Doutorado em Tecnologia Nuclear) / IPEN/T / Instituto de Pesquisas Energéticas e Nucleares - IPEN-CNEN/SP

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 master thesis deals with the facelift of the forecourt of the Dukovany Nuclear Power Station . This space consists of supporting functions for the main working such as administrative, metrology, stocks, cloakrooms and services. There is also the station for the buses, which carry employees to and from work, the regular bus station and parking lots. The forecourt of the power station is now inconvenient from a functional point of view, because the capacity of the existing buildings and parking lots is insufficient; and also from aesthetic point of view, because the buildings from 70’s don’t look good anymore, the parking consists of the large asphalt areas and there is no representative anteroom in front of the main entry, which should be there in view of the significance of the power station. In the study I deal with these problems by the complex reconstruction of the area, I replace the huge asphalt parking lots with the parking houses, create the administrative zone with the public place in front of the main entry to the plant, extend the capacities for the particular functions and add the new required functions (kindergaten, other services). The result of this conversion is the area with the particular functional zones with the representative forecourt in front of the main entry and with the enough space for each of the functions. This work follows up the last semester urbanist-architectonic study. The diploma thesis focus on the improvements of the weaker parts of the prime study and on the elaboration of the architectural study of the adminstrative zone with services.

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.