Public participation can be defined as ...”a process leading to a joint effort by stakeholders, technical specialists, the authorities and the proponent who work together to produce better decisions than if they had acted independently" (Greyling, 1999, p. 20). In South Africa, public participation processes are legally driven and form part a statutory part of environmental impact assessments. Many environmental impact assessments have been undertaken in South Africa, but the environmental impact assessment undertaken for the proposed construction of a demonstration module pebble bed modular reactor was perhaps one of the biggest studies undertaken to date from a public participation process point of view (Smit, 2003). The main aim of this mini-dissertation was to evaluate the public participation process followed for the environmental impact assessment of the demonstration module pebble bed modular reactor at Koeberg in the Western Cape Province against the criteria for fairness and competence as set out by Webler (In: Renn et al., 1995). Despite the fact that this work is eleven years old, it is still regarded as a benchmark for the evaluation of public participation processes in environmental decision making (Abelson et al., 2003). Webler (In: Renn et al., 1995) developed a normative theory for fairness and competence in public participation based on the theory of ideal speech of German sociologist Jürgen Habermas. Habermas’ main contribution to science was his theory of universal pragmatics (Author unknown, 2005). Universal pragmatics is a theory aimed at explaining how language is used to ensure mutual understanding and agreement. Webler (In: Renn et al., 1995) argues that the conditions of universal pragmatics, if applied to public participation, points towards the concepts of fairness (providing everyone with the opportunity to participate) and competence [providing participants (called interested and affected parties (I&APs) with the opportunity to make, question and validate speech acts]. Habermas advocates that each statement (or speech act) makes at least one validity claim and that there is a presupposition that the speaker can validate each claim to the satisfaction of all communication partners, should this be necessary (Perold, 2006). Furthermore, Habermas identifies four different types of validity claims, each having to do with a specific type of statement. In his theory, communicative speech acts have to do with comprehensibility; constantive speech acts with truth/correctness; regulative speech acts with normative rightness and representative speech acts with sincerity. Webler (In: Renn et al., 1995) developed a set of criteria to evaluate the fairness and competence in public participation. This set of criteria was applied to the public participation process of the case study. The study found that the process followed in the case study did not fare particularly well in either fairness or competence, but that fairness was slightly better than competence. The most alarming finding was that little attempt was made to ensure that validity claims – especially constantive (truth and factual information) – were validated or redeemed as this left the door open for misinterpretation, politics and incorrectness. It was also found that I&APs were, for the most, prevented from participating in the decision-making process. This finding may or may not be interpreted as negative as the public participation consultant never made a claim towards power sharing as well as the fact that there are widely differing opinions regarding the level to which public participation should take place. It was suggested that at least some elements of power sharing be incorporated into future processes, that validity claims – especially constantive (theoretical/factual) and therapeutic (regarding feelings and emotions) – must be able to stand up to scrutiny and should be validated. Finally, it was suggested that more attention be given to representative speech acts (statements regarding emotions, perceptions and feelings). / Dr. J. M. Meeuwis

The thesis deals with the communication of the Temelín Nuclear Power Plant. Nuclear sources of energy are currently a sensitive issue. The planned extension of the Temelín NPP is also a frequently discussed topic in the Czech Republic; it affects many stakeholders and faces increased public and media interest. The goal of the thesis is to create a communication strategy of the Temelín NPP targeted to the public. The draft of the strategy is based on the conclusions of the SWOT analysis focused on the Temelín NPP communication, furthermore, on the analysis of selected online communication tools used by the Temelín NPP, and mainly on the quantitative survey conducted with the purpose to find out the actual public perception of the Temelín NPP and the public attitude to nuclear energy in general. The strategy also uses the knowledge of the company ČEZ media analysis and the findings of the sociological survey among inhabitants of the area close to Temelín compiled by STEM/MARK research agency.

Um desafio recorrente em processos produtivos é o desenvolvimento de sistemas de monitoração e diagnóstico. Esses sistemas ajudam na detecção de mudanças inesperadas e interrupções, prevenindo perdas e mitigando riscos. Redes Neurais Artificiais (RNA) têm sido largamente utilizadas na criação de sistemas de monitoração. Normalmente as RNA utilizadas para resolver este tipo de problema são criadas levando-se em conta apenas parâmetros como o número de entradas, saídas e quantidade de neurônios nas camadas escondidas. Assim, as redes resultantes geralmente possuem uma configuração onde há uma total conexão entre os neurônios de uma camada e os da camada seguinte, sem que haja melhorias em sua topologia. Este trabalho utiliza o algoritmo de Otimização por Colônia de Formigas (OCF) para criar redes neurais otimizadas. O algoritmo de busca OCF utiliza a técnica de retropropagação de erros para otimizar a topologia da rede neural sugerindo as melhores conexões entre os neurônios. A RNA resultante foi aplicada para monitorar variáveis do reator de pesquisas IEA-R1 do IPEN. Os resultados obtidos mostram que o algoritmo desenvolvido é capaz de melhorar o desempenho do modelo que estima o valor de variáveis do reator. Em testes com diferentes números de neurônios na camada escondida, utilizando como comparativos o erro quadrático médio, o erro absoluto médio e o coeficiente de correlação, o desempenho da RNA otimizada foi igual ou superior ao da tradicional. / A recurring challenge in production processes is the development of monitoring and diagnosis systems. Those systems help on detecting unexpected changes and interruptions, preventing losses and mitigating risks. Artificial Neural Networks (ANN) have been extensively used in creating monitoring systems. Usually the ANN used to solve this kind of problem are created by taking into account only parameters as the number of inputs, outputs, and number of neurons in the hidden layers. This way, the result networks are generally fully connected and have no improvements in its topology. This work uses an Ant Colony Optimization (ACO) algorithm to create a tuned neural networks. The ACO search algorithm uses Back Error Propagation (BP) to optimize the network topology by suggesting the best neuron connections. The outcome ANN was applied to monitoring the IEA-R1 research reactor at IPEN. The results show that the algorithm is able to improve the performance of the model which estimates the values of the reactor variables. In tests with different numbers of neurons in the hidden layer, using as comparison the mean squared error, the mean absolute error, and the correlation coefficient, the performance of the optimized ANN proved equal or better than the equivalent traditional neural networks.

The topic of this work is the renewal of hermetic pipe penetrations within the primary loop of a nuclear power station. A description of the Dukovany nuclear power plant is included at the beginning, ranging from basic description to details of the renewed component and the reasoning for its replacement. Following is a description of the technologies, which are applied in order to restore the hermetic pipes, the choice of the placement of the heterogeneous weld and the compilation of a workflow for the renewal in the environment of the primary loop while maintaining the quality standards and the strict safety conditions. The conclusion includes a calculation of the expenses tied with the chosen variant the reasoning behind the renewal regarding the length of the outage.

The subject of this work is the repair of an air-handling unit. Possible causes of the problem are analyzed and multiple solutions are proposed. The first variant considers the refurbishment of the damaged component using thermal spray technology, including the choice of the best method. The second solution proposes the manufacture of a new part using machining in the environment of the company. This process requires an intermediate product to be made, and the choice of the correct machines from what the company has available, choice of the right tools and cutting conditions. The technical-economical evaluation compares the the expenses tied with both of the solutions. The chosen repair solution was used for the repair of the air-handling unit.

The goal of this thesis is to simulate the course of events during Dukovany Nuclear Power Plant blackout and to determine the optimal process of cooling down the cold branches of the primary circuit loops to secure the maximum amount of the primary fuel needed for the residual heat outlet so that the operating staff has as much time as possible for renewing the electric power supply. The first part of the thesis describes nuclear power plants built in the Czech Republic and in the world as well as reactor blocks whose construction is being considered. The detailed description of Dukovany’s steam generator is shared in chapter three. Following chapter summarizes blackouts that occurred at power plants around the world, events that might have led to blackouts in the Czech Republic, and it also describes current approach to blackout problematics at Dukovany Nuclear Power Plant. Chapters six and seven contain the core of the thesis. That includes detailed description of a mathematical model explaining the behaviour of a reactor block during blackout as well as the analysis of the results found.

The master's thesis deals with the analysis of source term of the spent nuclear fuel of the Dukovany Nuclear Power Plant in order to determine the proposals for the transfer of spent nuclear fuel to a deep geological repository in the Czech Republic. To introduce the reader into the issue are briefly described the main aspects, such as the development of the nuclear fuel used in the history of the Dukovany Nuclear Power Plant. These aspects have an influence on the final draft of the timetable. One of the important partial tasks is the processing of an estimate of the future range of spent nuclear fuel, which is based on the current ideas of company ČEZ, a.s. for the future direction of the fuel cycle at the Dukovany nuclear power plant. For the purposes of this work, the key data are the time dependencies of radioactivity and the development of residual heat in individual fuel assemblies. This data are calculated by the software called PAL440_R4, based on the prepared estimates of the spent nuclear fuel assortment. The calculated data are then edited and sorted by MS Excel. For the sake of completeness, the characteristic values and the time dependencies of the radioactivity and the development of residual heat in fuel assemblies. Final timetables for the transfer of spent nuclear fuel to a deep geological repository are processed in several variants, and their selection and application options are justified. For illustration are important parameters given in the form of tables and charts.

Viscous dampers are widely used to ensure seismic resistance of pipelines and equipment in nuclear power plants. Damping characteristics of these dampers are nonlinearly frequency dependent and thus causing complications in computational modelling of seismic response. Engineers commonly use two ways to deal with this nonlinearity: The first option is to consider damper by means of “snubber”. This is essentially linear spring element that is active for dynamic load and does not resist static loads. Snubber behaviour during seismic event is described by a equivalent stiffness (sometimes called pseudostiffness). The equivalent stiffness could be defined by the iterative calculations of piping natural frequencies and mode shapes taking into account seismic excitation. However, in complicated structures such as the main circulation loop of nuclear power plant the iterative calculation is difficult and could bring significant inaccuracies. On the other hand, the benefit of such modelling is a possibility to apply the commonly used linear response spectrum method for a solution. The second option is to describe damping characteristics using suitable rheological model. The seismic response is than determined by direct integration of the equations of motion. The behaviour of dampers is described exactly enough but the calculation and post-processing, especially nodal stresses time-histories, are time consuming. The goal of this work was to find a methodology for determining the seismic response of complex pipe systems with viscous dampers. Methodology allows a sufficiently accurate determination of the seismic response of piping systems and also allows obtaining of the results in effective time. The procedure is as follows. Firstly, specialized piping program (AutoPIPE) is used for the development of computational model. Next step is to determine a static response of structure and its verification with experimental measurements, if possible. Using script in Python language a computational model is converted from AutoPIPE into general finite element model in ANSYS system. Four-parameter Maxwell rheological model is used to describe behaviour of viscous dampers. Seismic load is represented by synthetic accelerograms. Newmark algorithm of direct integration of the equation of motion is used to obtain seismic response (only reactions and displacements in nodes of interest are necessary). Than is the equivalent stiffness is than gained from displacements and reactions as median value of their ratios. Received stiffness are subsequently transferred to AutoPIPE program where the seismic solution is performed using response spectra method. Finally, the dynamic response is combined with the static response and stress assessment according standards is done. The created methodology was applied in the seismic resistance calculation of the main circulation piping and piping of pressurizer in nuclear power plants type VVER 440 and type VVER 1000.

This thesis is focused on the design of passive cooling system for a nuclear reactor VVER- 1000.This type of reactor is located in the Czech Republic in the location of Nuclear power plant Temelín. The thesis states an overview of the different cooling systems of nuclear power plants. The thesis is focused on passive safety system especially on passive cooling system, so there was done an overview of currently used passive safety system. In the work is discussed nuclear safety and the maximum projected accident of VVER-1000, which is called LOCA accident. In the design part of the thesis was done thermal calculation of heat exchangers. Exchangers are designed as condensers with a natural flow, where cooling of system is provided by outside airflow in case an accident. The results are evaluated at the end of the thesis.

The flow distribution in the primary circuit of the pressurized water reactor was studied with experiments and Computational Fluid Dynamics (CFD) simulations. The main focus was on the flow field and mixing in the downcomer of the pressure vessel: how the different factors like the orientation of operating loops, the total loop flow rate and the asymmetry of the loop flow rates affect the outcome. In addition to the flow field studies the overall applicability of CFD methods for primary circuit thermal-hydraulic analysis was evaluated based on the CFD simulations of the mixing experiments of the ROCOM (Rossendorf Coolant Mixing Model) test facility and the mixing experiments of the Paks NPP. The experimental part of the work in work package 3 included series of steady state mixing experiments with the ROCOM test facility and the publication of results of Paks VVER-440 NPP thermal mixing experiments. The ROCOM test facility models a 4-loop KONVOI type reactor. In the steady-state mixing experiments the velocity field in the downcomer was measured using laser Doppler anemometry and the concentration of the tracer solution fed from one loop was measured at the downcomer and at the core inlet plane. The varied parameters were the number and orientation of the operating loops, the total flow rate and the (asymmetric) flow rate of individual loops. The Paks NPP thermal mixing experiments took place during commissioning tests of replaced steam generator safety valves in 1987-1989. It was assumed that in the reactor vessels of Paks VVER-440 NPP equipped with six loops the mixing of the coolant is not ideal. For the realistic determination of the active core inlet temperature field for the transients and accidents associated with different level temperature asymmetry a set of mixing factors were determined. Based on data from the online core monitoring system and a separate mathematical model the mixing factors for loop flows at the core inlet were determined. In the numerical simulation part of the work package 3 the detailed measurements of ROCOM tests were used for the validation of CFD methods for primary circuit studies. The selected steady state mixing experiments were simulated with CFD codes CFX-4, CFX-5 and FLUENT. The velocity field in the downcomer and the mixing of the scalar were compared between CFD simulations and experiments. The CFD simulations of full scale PWR included the simulation of Paks VVER-440 mixing experiment and the simulation of Loviisa VVER-440 downcomer flow field. In the simulations of Paks experiments the experimental and simulated concentration field at the core inlet were compared and conclusions made concerning the results overall and the VVER-440 specific geometry modelling aspects like how to model the perforated elliptic bottom plate and what is the effect of the cold leg bends to the flow field entering to the downcomer. With Loviisa simulations the qualitative comparison was made against the original commissioning experiments but the emphasis was on the CFD method validation and testing. The overall conclusion concerning the CFD modelling of the flow field and mixing in the PWR primary circuit could be that the current computation capacity and physical models also in commercial codes is beginning to be sufficient for simulations giving reliable and useful results for many real primary circuit applications. However the misuse of CFD methods is easy, and the general as well as the nuclear power specific modelling guidelines should be followed when the CFD simulations are made.