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.

This diploma thesis focuses on strength design of steam condenser. The goal of the thesis is to make strength calculations for the specific operation conditions, introduce possible solutions, provide recommendations and refer to weak points of such calculation procedures. First, thermal-hydraulic design in HTRI software is performed. Strength calculations respect ČSN EN 13445 standard. Strength calculation with imported temperature field is performed in ANSYS Workbench software. In the next step, another strength calculation is realized in Sant´ Ambrogio software. Results are evaluated in conclusion chapter, including recommendations for the possible following work.

The master thesis is focused on analysis of specific cross-flow in-line tube bundle heat exchanger which deals with significant operational problems. Thermal, hydraulic and vibration calculation analysis of selected parts of the heat exchanger is performed based on CFD flow distribution results. Calculation is performed by using Maple software. Thermal and hydraulic calculations are compared with results obtained by commercial software HTRI.

A breed-and-burn molten salt reactor (BBMSR) concept is proposed to address the Generation IV fuel cycle sustainability objective in a once-through cycle with low enrichment and no reprocessing. The BBMSR uses separate fuel and coolant molten salts, with the fuel contained in assemblies of individual tubes that can be shuffled and reclad periodically to enable high burnup. In this dual-salt configuration, the BBMSR may overcome several limitations of previous breed-and-burn (B$\&$B) designs to achieve high uranium utilisation with a simple, passively safe design. A central challenge in design of the BBMSR fuel is balancing the neutronic requirement of large fuel volume fraction for B$\&$B mode with the thermal-hydraulic requirements for safe and economically competitive reactor operation. Natural convection of liquid fuel within the tubes aids heat transfer to the coolant, and a systematic approach is developed to efficiently model this complex effect. Computational fluid dynamics modelling is performed to characterise the unique physics of the system and produce a new heat transfer correlation, which is used alongside established correlations in a numerical model. A design framework is built around this numerical model to iteratively search for the limiting power density of a given fuel and channel geometry, applying several defined temperature and operational constraints. It is found that the trade-offs between power density, core pressure drop, and pumping power are lessened by directing the flow of coolant downwards through the channel. Fuel configurations that satisfy both neutronic and thermal-hydraulic objectives are identified for natural, 5$\%$ enriched, and 20$\%$ enriched uranium feed fuel. B$\&$B operation is achievable in the natural and 5$\%$ enriched versions, with power densities of 73 W/cm$^3$ and 86 W/cm$^3$, and theoretical uranium utilisations of 300 $\mathrm{MWd/kgU_{NAT}}$ and 25.5 $\mathrm{MWd/kgU_{NAT}}$, respectively. Using 20$\%$ enriched feed fuel relaxes neutronic constraints so a wider range of fuel configurations can be considered, but there is a strong inverse correlation between power density and uranium utilisation. The fuel design study demonstrates the flexibility of the BBMSR concept to operate along a spectrum of modes ranging from high fuel utilisation at moderate power density using natural uranium feed fuel, to high power density and moderate utilisation using 20$\%$ uranium enrichment.

La thèse porte sur la simulation de la thermohydraulique et des transferts thermiques dans un faisceau d'aiguilles d'assemblage combustible de réacteur à neutrons rapides à caloporteur sodium.Des premiers calculs ont été réalisés par une approche moyennée de type RANS à l'aide du code industriel STAR-CCM+. De cette modélisation, il ressort une meilleure compréhension des transferts de chaleur opérés entre les aiguilles et le sodium. Les principales grandeurs macroscopiques de l'écoulement sont en accord avec les corrélations. Cependant, afin d'obtenir une description détaillée des fluctuations de température au niveau des fils espaceur, une approche plus détaillée de type LES et DNS est apparue indispensable. Pour la partie LES, le code TRIO_U a été utilisé. Concernant la partie DNS, un code de recherche a été utilisé. Ces approches requièrent des temps de calculs considérables qui ont nécessité des géométries représentatives mais simplifiées.L'approche DNS permet d'étudier l'écoulement à bas nombre de Prandtl, qui induit un comportement très différent du champ thermique relativement au champ hydraulique. Le calcul LES de l'assemblage montre que la présence du fil espaceur génère l'apparition de points chauds locaux (~20°C) en aval de celui-ci par rapport à l'écoulement sodium, au niveau de son contact avec l'aiguille. Les fluctuations de température au niveau des fils espaceur sont faibles (~1°C-2°C). En régime nominal, l'analyse spectrale montre l'absence de grande amplitude d'oscillations de température à basse fréquence (2-10 Hz); les conséquences sur la tenue mécanique des structures devront être analysées. / The thesis focuses on the numerical simulation of sodium flow in wire wrapped sub-assembly of Sodium-cooled Fast Reactor (SFR).First calculations were carried out by a time averaging approach called RANS (Reynolds- Averaged Navier-Stokes equations) using industrial code STAR-CCM+. This study gives a clear understanding of heat transfer between the fuel pin and sodium. The main variables of the macroscopic flow are in agreement with correlations used hitherto. However, to obtain a detailed description of temperature fluctuations around the spacer wire, more accurate approaches like LES (Large Eddy Simulation) and DNS (Direct Numerical Simulation) are clearly needed. For LES approach, the code TRIO_U was used and for the DNS approach, a research code was used. These approaches require a considerable long calculation time which leads to the need of representative but simplified geometry.The DNS approach enables us to study the thermal hydraulics of sodium that has very low Prandtl number inducing a very different behavior of thermal field in comparison to the hydraulic field. The LES approach is used to study the local region of sub-assembly. This study shows that spacer wire generates the local hot spots (~20°C) on the wake side of spacer wire with respect to the sodium flow at the region of contact with the fuel pin. Temperature fluctuations around the spacer wire are low (~1-2°C). Under nominal operation, the spectral analysis shows the absence of any dominant peak for temperature oscillations at low frequency (2-10Hz). The obtained spectra of temperature oscillations can be used as an input for further mechanical studies to determine its impact on the solid structures.

The master thesis deals with thermal hydraulic design and strength design of a heat exchanger with “U” tube bundle inside of a shell. The first chapter introduces general design issues of the heat exchangers. The following chapter describes thermal hydraulic design created in software Maple 16.0 by using Kern’s method and the method of Bell-Delaware. HTRI software was used for the control of thermal hydraulic design correctness. To check critical locations of fluid flow in space between the tubes, the CFD model was created at ANSYS Fluent 14.0 software. Accuracy of strength design was verifying by Sant’ Ambrogio software in accordance with ČSN EN 13 445 standards. The last chapter concerns with FEM analysis. According to standards ČSN EN 13 445 the design by analysis namely method based on stress categories were used for the strength analysis of nozzle.

This thesis is focused on the thermal-hydraulic and construction design of heat exchanger with floating head. The introductory part is dedicated to the design of heat exchangers. Next part is focused on the thermal-hydraulic design created in HTRI software (module Xist. The next section is the strength calculation of selected components of the heat exchanger according to EN 13445 in software Sant'Ambrogio. The following part describes each of the analysis in software Ansys Workbench. The final part contains complete manufacturing documentation of heat exchanger.

Nowadays, one of the solutions considered in order to face the issue of global warming and to move towards a carbon neutral society relies on the use of solar energy as a renewable and bountiful primary source. And, if photovoltaic technologies account for a large part in the solar energy market, recent years have witnessed the growth of non-concentrated and concentrated solar thermal technologies. Among them, concentrated solar power technology (CSP) which uses the optical concentration of direct solar irradiation to generate high pressure and high temperature steam in the absorber tubes of the plant, has become a promising approach reaching 4.9 GWe of installed capacity by the end of 2015 [1]. However, one of the main challenges faced by CSP technology concerns the variability of solar energy related for example to sunrise, sunset, passing clouds… In addition to that, when it comes to direct steam generation, the presence of a two-phase flow regime inside the absorber tubes leads to a strong dynamic behavior of the steam generation. It is consequently necessary to be able to simulate this dynamic behavior in order to better handle the design and operation of CSP plants. Such simulation tools can then be used for the implementation and the test of reliable control systems aimed at maintaining desired operating conditions in spite of changes in solar irradiation. In this context, the National Institute for Solar Energy (INES), part of the French Alternative Energies and Atomic Energy Commission (CEA) wishes to upgrade their dynamic simulation tool that would enable its teams to reproduce the behavior of a prototype based on the Fresnel solar field technology including direct steam generation which was built and commissioned at Cadarache, Aix-en-Provence. This Master thesis work takes place within this framework and aims at developing a multi-domain dynamic model of the aforementioned prototype. To do so, three models respectively in the thermalhydraulic, the optical and the control-command domains are built and combined using a co-simulation approach relying on an in-house simulation platform called PEGASE. More specifically the development of the following models has been addressed:  a thermal-hydraulic model of the two-phase flow circulating inside the vaporizer field of the prototype and realized with the thermal-hydraulic code CATHARE [2] (Advanced ThermalHydraulic Code for Water Reactor Accidents) applied to solar thermal biphasic issues,  an optical model of the receiver programmed using the Modelica language and the Dymola (Dynamic Modelling Laboratory) simulation software,  control-command models (PID controller, control architecture…) adapted and built upon blocks taken from a modelling library included in the PEGASE platform. Each model was first developed and tested on a standalone basis. These models were then coupled using the PEGASE co-simulation platform. A sunny day was simulated using the multi-domain model and the controllability of the plant was analyzed. At this stage, the study focused on the steam separator level regulation. A thermal-hydraulic study also focused on potential instabilities in the vaporizer that can occur under certain circumstances of water temperature at vaporizer inlet and solar heat flux. This analysis was carried out with a CATHARE standalone model. Perspectives of the present work include a complete validation of the developed models from future experimental data and further developments should aim to extend the modelling scope of the numerical simulator towards a representation of all the hydraulic parts of the CSP prototype. Control schemes and regulation tools would have to be extended as well in order to move towards a more representative control architecture of the prototype. Particularly, the steam quality at vaporizer outlet is an important variable to regulate. Indeed, this parameter is usually kept between 60% and 80% [3]. It must be high enough to limit the power consumption of the recirculation pump but not too high in order to prevent absorber dry-out. / Solenergi, som är en förnybar och riklig primärkälla, är en av de lösningarna som anses kunna lösa problemet med global uppvärmning och bidrar i omvandlingen till ett kolneutralt samhälle. Andelen fotovoltaiska teknologier på energimarknaden är övervägande, men andelen koncentrerad och ickekoncentrerad solterminsteknik har ökat under de senaste åren. Bland solterminsteknikerna är koncentrerad solenergiteknik (CSP), som använder den optiska koncentrationen av direkt strålning för att generera högtrycks- och högtemperaturånga i anläggningens absorberarrör, ett lovande tillvägagångssätt som har nått 4.9 GWe installerad kapacitet i slutet av 2015 [1]. En av de största utmaningarna med CSP-tekniken är solenergins variation vid till exempel soluppgång, solnedgång och passerande moln, vilket beror på varierad tillgång av solljus. Det finns också utmaningar med direkt ånggenerering via tvåfasflödes regimer inuti absorberarrören eftersom det leder till ett starkt dynamiskt beteende vid ånggenereringen. Det är följaktligen nödvändigt att kunna simulera detta dynamiska beteende för att bättre hantera design och drift av CSP-anläggningar. Sådana simuleringsverktyg kan sedan användas för att genomföra tester för att erhålla tillförlitliga styrsystem som upprätthåller önskade driftsförhållanden trots förändringar i solstrålningen I detta sammanhang vill National Institute for Solar Energy (INES), som är en del av den franska alternativa energikommissionen och atomenergi kommissionen (CEA), förbättra dess dynamiskt simuleringsverktyg som skulle möjliggöra för sina team att reproducera beteendet hos en prototyp baserad på Fresnel solfältsteknik inklusive direkt ånggenerering som byggts och beställts vid Cadarache, Aix-enProvence. Denna masteruppsats sker inom ramen för detta och syftar till att utveckla en dynamisk modell med flera domäner av den ovan nämnda prototypen. Tre modeller i termisk-hydraulisk, optisk och kontrollkommando domäner har byggts och kombinerats med hjälp av en co-simuleringsmetod som bygger på en intern simuleringsplattform som heter PEGASE. Mer specifikt om utvecklingen av modellerna enligt nedan:  En termisk-hydraulisk modell av tvåfasflöde som cirkulerar inuti förångarens fält på prototypen har realiserats med termisk-hydraulisk kod CATHARE [2] (Advanced Thermal-Hydraulic Code for Water Reactor Accidents) som appliceras på soltermisk bifasiska frågeställningar.  En optisk modell av mottagaren har programmerats med hjälp av Modelica-språket och simuleringsprogrammet Dymola (Dynamic Modeling Laboratory).  Modeller av kontrollkommandon (PID-kontroller, kontrollarkitektur ...) har byggts och anpassats i moduler som hämtats från modelleringsbibliotek som ingår i PEGASE-plattformen. Varje modell utvecklades och testades på fristående basis. Modellerna kopplades sedan samman i PEGASE-co-simuleringsplattformen. En solig dag simulerades därefter med en flerdomänmodell och styrningsförmågan av anläggningen analyserades. Vid detta stadium fokuserade studien på att reglera nivån av ångseparerande. En termisk-hydraulisk studie fokuserade sedan på potentiella instabiliteter i förångaren som kan uppstå under vissa omständigheter av vatteninloppstemperatur och solvärmeflöde. Denna analys genomfördes med en CATHARE fristående modell. Perspektiven för det aktuella arbetet omfattar en fullständig validering av de utvecklade modellerna med hjälp av framtida experimentella data. Vid en vidareutveckling bör inriktningen vara att utvidga modellernas omfattning av den numeriska simulatorn till att representera alla hydrauliska delar av CSP prototypen. Styrsystem och regleringsverktyg skulle också behöva förbättras för att få en mer representativ kontroll arkitektur av prototypen. I synnerhet är ångkvaliteten vid förångarens utlopp en viktig variabel att reglera. Faktum är att den här parametern vanligtvis hålls mellan 60% och 80% [3]. Det måste vara tillräckligt högt för att begränsa recirkulationspumpens elförbrukning men inte för hög för att förhindra att absorberen torkar ut.

The nuclear industry has long relied upon bounding parametric analyses in predicting the safety margins of reactor designs undergoing design-basis accidents. These methods have been known to return highly-conservative results, limiting the operating conditions of the reactor. The Best-Estimate Plus Uncertainty (BEPU) method using a modernized version of the Code-Scaling, Applicability, and Uncertainty (CSAU) methodology has been applied to more accurately predict the safety margins of the Oregon State University Advanced Plant Experiment (APEX) facility experiencing a Loss-of-Feedwater Accident (LOFA). The statistical advantages of the Bayesian paradigm of probability was utilized to incorporate prior knowledge when determining the analysis required to justify the safety margins. RELAP5 Mod 3.3 was used to accurately predict the thermal-hydraulics of a primary Feed-and-Bleed response to the accident using assumptions to accompany the lumped-parameter calculation approach. A novel coupling of thermal-hydraulic and statistical software was accomplished using the Symbolic Nuclear Analysis Package (SNAP). Uncertainty in Peak Cladding Temperature (PCT) was calculated at the 95/95 probability/confidence levels under a series of four separate sensitivity studies. / Graduation date: 2013