This study identifies two syntactically distinguishable types of Serial Verb Constructions (SVCs) in Mandarin Chinese (MC) and Jinjiang Southern Min (JSM), corresponding to the nuclear and core distinction made in Role and Reference Grammar (Foley and Van Valin 1984, Foley and Olson 1985, Van Valin and LaPolla 1997). This distinction is also made on the basis of a general consensus of the cross-linguistic classifications of the processes of monoclausal multi-verb construction formation (e.g., Butt 1993, 1997, Baker and Harvey 2010): namely, predicate fusion and argument fusion. In this study, I propose two sets of diagnostics to establish the distinction; these go beyond the range covered in previous studies (e.g., Olson 1981, Foley and Olson 1985, Crowley 2002, Chang 2007). In the first set of diagnostics in this study, seven inter-clausal diagnostics are considered as the threshold where the behaviours of bi-clausal structures and SVCs split. These diagnostics include independent negation, passivisation of the object of V2, independent modification by temporal adverbial, independent marking of viewpoint aspect, independent modification by manner adverbial, prosodic structure and the Coordinate Structure Constraint (Ross 1967) that is employed in a more restricted manner. In the second set of diagnostics, four intra-clausal diagnostics are adopted to make the distinction between nuclear and core SVCs, which include passivisation of O1, insertion of intervening material, coordination within the SVC, and obligatory topicalisation of undergoer argument. Of particular interest is the possibility that the same string of verbs may occur in superficially similar, but structurally different, SVCs: for example, the Cause-Effect SVC and the Excessive SVC. The diagnostics employed in this study are proposed as a novel method to establish the distinction between the SVCs and the bi-clausal structures, and more importantly, between core and nuclear types of SVC. Contributing to the originality of the new method of diagnosing the status of the SVCs proposed in this study, I add five novel diagnostics, such as passivisation of the object of V2, independent marking of viewpoint aspect, tone sandhi between adjacent verbs, coordination within the SVC and obligatory topicalisation of the undergoer argument, in addition to those that have been employed in the literature. I restrict myself to data of MC and JSM in discussing the rationale of the diagnostics. However, this novel method of identifying SVCs is expected to be cross-linguistically applicable with consistent results, while at the same time allowing for the possibility of cross-linguistic differences in the semantic sub-types of SVCs identified in each language.

Westinghouse Electric Sweden AB’s three-dimensional reactor core simulation program POLCA uses a large number of tables containing various fuel dependent data, such as cross sections, pin power maps with power distribution etc. POLCA uses quadratic and linear interpolation to extract the values needed for the simulation. However, finding the correct values to interpolate between takes time. This master thesis describes a method of compacting the tables by removing values, in order to shorten the needed simulation time. This is done so that no significant accuracy is lost in the interpolations. The method also finds deviant values and replaces these with new, interpolated values. The thesis shows that approximately 90 % of all values can be removed without losing significant accuracy. These results are however heavily dependent on the choice of accuracy loss criterion; a lower allowance for accuracy loss lowers the amount of values which can be removed sharply.

Pour la conception des cœurs de réacteurs de 4ème génération, une précision accrue est requise pour les calculs des différents paramètres neutroniques. Les ressources mémoire et le temps de calcul étant limités, une solution consiste à utiliser des méthodes de raffinement de maillage afin de résoudre l'équation de transport des neutrons. Le flux neutronique, solution de cette équation, dépend de l'énergie, l'angle et l'espace. Les différentes variables sont discrétisées de manière successive. L'énergie avec une approche multigroupe, considérant les différentes grandeurs constantes sur chaque groupe, l'angle par une méthode de collocation, dite approximation Sn. Après discrétisation énergétique et angulaire, un système d'équations hyperboliques couplées ne dépendant plus que de la variable d'espace doit être résolu. Des éléments finis discontinus sont alors utilisés afin de permettre la mise en place de méthodes de raffinement dite hp. La précision de la solution peut alors être améliorée via un raffinement en espace (h-raffinement), consistant à subdiviser une cellule en sous-cellules, ou en ordre (p-raffinement) en augmentant l'ordre de la base de polynômes utilisée.Dans cette thèse, les propriétés de ces méthodes sont analysées et montrent l'importance de la régularité de la solution dans le choix du type de raffinement. Ainsi deux estimateurs d'erreurs permettant de mener le raffinement ont été utilisés. Le premier, suppose des hypothèses de régularité très fortes (solution analytique) alors que le second utilise seulement le fait que la solution est à variations bornées. La comparaison de ces deux estimateurs est faite sur des benchmarks dont on connaît la solution exacte grâce à des méthodes de solutions manufacturées. On peut ainsi analyser le comportement des estimateurs au regard de la régularité de la solution. Grâce à cette étude, une stratégie de raffinement hp utilisant ces deux estimateurs est proposée et comparée à d'autres méthodes rencontrées dans la littérature. L'ensemble des comparaisons est réalisé tant sur des cas simplifiés où l'on connaît la solution exacte que sur des cas réalistes issus de la physique des réacteurs.Ces méthodes adaptatives permettent de réduire considérablement l'empreinte mémoire et le temps de calcul. Afin d'essayer d'améliorer encore ces deux aspects, on propose d'utiliser des maillages différents par groupe d'énergie. En effet, l'allure spatiale du flux étant très dépendante du domaine énergétique, il n'y a a priori aucune raison d'utiliser la même décomposition spatiale. Une telle approche nous oblige à modifier les estimateurs initiaux afin de prendre en compte le couplage entre les différentes énergies. L'étude de ce couplage est réalisé de manière théorique et des solutions numériques sont proposées puis testées. / The different neutronic parameters have to be calculated with a higher accuracy in order to design the 4th generation reactor cores. As memory storage and computation time are limited, adaptive methods are a solution to solve the neutron transport equation. The neutronic flux, solution of this equation, depends on the energy, angle and space. The different variables are successively discretized. The energy with a multigroup approach, considering the different quantities to be constant on each group, the angle by a collocation method called Sn approximation. Once the energy and angle variable are discretized, a system of spatially-dependent hyperbolic equations has to be solved. Discontinuous finite elements are used to make possible the development of $hp-$refinement methods. Thus, the accuracy of the solution can be improved by spatial refinement (h-refinement), consisting into subdividing a cell into subcells, or by order refinement (p-refinement), by increasing the order of the polynomial basis.In this thesis, the properties of this methods are analyzed showing the importance of the regularity of the solution to choose the type of refinement. Thus, two error estimators are used to lead the refinement process. Whereas the first one requires high regularity hypothesis (analytical solution), the second one supposes only the minimal hypothesis required for the solution to exist. The comparison of both estimators is done on benchmarks where the analytic solution is known by the method of manufactured solutions. Thus, the behaviour of the solution as a regard of the regularity can be studied. It leads to a hp-refinement method using the two estimators. Then, a comparison is done with other existing methods on simplified but also realistic benchmarks coming from nuclear cores.These adaptive methods considerably reduces the computational cost and memory footprint. To further improve these two points, an approach with energy-dependent meshes is proposed. Actually, as the flux behaviour is very different depending on the energy, there is no reason to use the same spatial discretization. Such an approach implies to modify the initial estimators in order to take into account the coupling between groups. This study is done from a theoretical as well as from a numerical point of view.

Dans cette thèse, nous nous intéressons à la résolution des équations SPN du transport de neutrons au sein des cœurs de réacteurs nucléaires à eau pressurisée. Ces équations forment un problème aux valeurs propres généralisé. Dans notre étude nous commençons par le problème source associé et ensuite nous étudions le problème aux valeurs propres. Un cœur de réacteur est composé de différents milieux: le combustible, le fluide caloporteur, le modérateur... à cause de ces hétérogénéités de la géométrie, le flux solution du problème source peut être peu régulier. Nous proposons l’analyse numérique de l’approximation de la solution par la méthode des éléments finis du problème source dans le cas où la solution est peu régulière. Pour le problème aux valeurs propres, dans le cas mixte, les théories déjà développées ne s’appliquent pas. Nous proposons ici une nouvelle méthode pour étudier la convergence de la méthode des éléments finis mixtes pour les problèmes aux valeurs propres. Pour les solutions peu régulières, la montée en ordre de la méthode des éléments finis n’améliore pas l’approximation du problème, il faut raffiner le maillage aux alentours des singularités de la solution. La géométrie des cœurs de réacteur se prête bien aux maillages cartésiens, mais leur raffinement augmente vite leur nombre de degrés de liberté. Pour palier à cette augmentation, nous proposons ici une méthode de décomposition de domaine qui permet d’utiliser des maillages globalement non-conformes. / In this thesis, we investigate the resolution of the SPN neutron transport equations in pressurized water nuclear reactor. These equations are a generalized eigenvalue problem. In our study, we first considerate the associated source problem and after we concentrate on the eigenvalue problem. A nuclear reactor core is composed of different media: the fuel, the coolant, the neutron moderator... Due to these heterogeneities of the geometry, the solution flux can have a low-regularity. We propose the numerical analysis of its approximation with finite element method for the low regular case. For the eigenvalue problem under its mixed form, we can not rely on the theories already developed. We propose here a new method for studying the convergence of the SPN neutron transport eigenvalue problem approximation with mixed finite element. When the solution has low-regularity, increasing the order of the method does not improve the approximation, the triangulation need to be refined near the singularities of the solution. Nuclear reactor cores are well-suited for Cartesian grids, but the refinement of these sort of triangulations increases rapidly their number of degrees of freedom. To avoid this drawback, we propose domain decomposition method which can handle globally non-conforming triangulations.

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.

Devant l’augmentation de la part des énergies renouvelables en France, cette thèse propose d’étudier l’augmentation de la flexibilité des réacteurs à eau pressurisée en croisant deux disciplines pour, chacune, atteindre des objectifs complémentaires : l’Ingénierie Système (IS) et l’Automatique.Dans le contexte de l’ingénierie de systèmes complexes et du Model Based Systems Engineering, ce travail propose dans un premier temps une méthode de conception se fondant sur les principes normatifs de l’IS et respectant les habitudes et les pratiques courantes en ingénierie de Framatome. Cette méthode a pour vocation de formaliser et assurer le passage des exigences aux architectures et d’améliorer les capacités de vérification des modèles développés lors de la conception. Elle s’organise autour de langages de modélisation interopérables, couvrant l’ensemble des processus promus par l’IS. La méthode proposée est appliquée sur le système dont les performances sont les plus limitantes dans le contexte de l’augmentation de flexibilité : le Core Control. Ce composant algorithmique du réacteur assure le contrôle des paramètres de fonctionnement du cœur : la température moyenne, la distribution axiale de puissance et la position des groupes de grappes.La thèse propose ensuite des contributions techniques relevant du champ de l’Automatique. Il s’agit de concevoir un système de régulation répondant aux exigences issues de la formalisation IS évoquée ci-dessus. La solution proposée repose sur une stratégie de commande hiérarchisée, utilisant la complémentarité des approches dites de commande multi-objectif, de séquencement de gains et enfin de commande prédictive. Un modèle de réacteur nucléaire simplifié innovant est développé à des fins de conception du système de régulation et de simulations intermédiaires. Les résultats obtenus ont montré les capacités d’adaptation de la démarche proposée à des spécifications diverses. Les performances atteintes sont très encourageantes lorsque évaluées en simulation à partir d’un modèle réaliste et comparées à celles obtenues par les modes de pilotages classiques. / In the event of increasing renewable energies in France, this thesis proposes to study the flexibility increase of pressurized water reactors (PWR) throughout two different engineering disciplines aiming at complementary objectives: Systems Engineering (SE) and Control theory.In a first phase, within the frame of complex systems design and Model Based Systems Engineering, this work proposes a SE method based on SE standard principles and compliant with Framatome’s practices and addressing the revealed issues. This SE contribution is twofold: formalize and ensure the path from requirements to system architectures and enhance the capabilities of models verification. The method revolves around interoperable modeling languages, covering the SE processes: from requirement engineering to system architecture design. The method is applied to the system, which performances are the most limiting in the context of flexibility increase: the Core Control. This algorithmic reactor component ensures the control of: the average coolant temperature, the axial offset and the rod bank position, three of the core main functioning parameters.In order to provide a technical contribution relying on some advanced control methodologies. It consists in designing a control system meeting the requirements defined by the SE method application. The proposed solution is in a two-layer control strategy using the synergies of multi-objective control, gain-scheduling and predictive control strategies. A simplified innovative nuclear reactor model is employed to conceive the control algorithm, simulate and verify the developed models. The results obtained from this original approach showed the ability to adapt to various specifications. Compared to conventional core control modes, the simulation results showed very promising performances, while meeting the requirements, when evaluated on a realistic reactor model.

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.