Thesis (Ph.D.)--Massachusetts Institute of Technology, Dept. of Physics, 1980. / Vita. / Includes bibliographical references. / by Alan Charles Janos. / Thesis (Ph.D.)--Massachusetts Institute of Technology, Dept. of Physics, 1980.

This thesis reports observations of kink and tearing modes in the High Beta Tokamak - ExtendedPulse (HBT-EP) experiment. When unstable, these modes could limit the operation of tokamaks used for fusion power by terminating the plasma discharge and causing rapid loss of plasma energy. The aim of this work is to characterize the sudden transition after a sawtooth crash of coupled 2/1-3/1 tearing-kink modes into a sustained and disruptive 2/1 tearing mode. The following diagnostic techniques are used. Kink and tearing modes in HBT-EP distort the plasma edge, measured by a large array of Mirnov sensors, and perturb the interior of the plasma, observed routinely with Extreme Ultraviolet (EUV) detector arrays. Two arrays, with different transmission filters, are located with tangential views to estimate the time evolution of the plasma temperature profile. Four EUV arrays, with 16 detectors each, are positioned with different poloidal views for poloidal Extreme Ultraviolet (pEUV) emission tomography. The 2D emissive structures producing the pEUV signals are reconstructed with tomographic inversion using a pixel basis and fixed weighting smoothness regularization. Spatial and temporal correlations across these independent diagnostics are used to measure the evolution and structure of coupled modes using a technique called multidiagostic Singular Value Decomposition (mdSVD). In mdSVD, orthogonal modes are identified within any fixed time window with their unique spatial and temporal characteristics. The technique uncovers: coherent behavior of coupled (𝑚/𝑛) = (2/1) and (3/1) tearing-kink modes and rapid changes in plasma structure associated with sawtooth crashes which trigger disruptive and nondisruptive tearing modes. HBT-EP’s unique radially movable wall is found to significantly influence sawtooth triggering of disruptive tearing modes. The onset of sawtooth-triggered modes depends both on the plasma-wall separation, or wall coupling, and on the value of edge safety factor qₐ. We confirm that the condition for sawtooth triggering of disruptive (𝑚/𝑛) = (2/1) tearing modes does not correspond to the mode’s single-helicity stability condition Δ′₂/₁. We identify a dependency of the sawtooth period 𝝉_𝑠𝑡 on the wall position and qa as a candidate to explain the onset of the saturated tearing mode. This thesis motivates future efforts to model the influence of a nearby resistive wall on sawtooth triggering of tearing modes.

In this work we study the magnetic field modeling of realistic non-axisymmetric plasma equilibrium configurations and the heat flux patterns on the plasma facing components of tokamak divertor discharges. We start by establishing the relation between generic magnetic configurations and Hamiltonian dynamical systems. We apply the concept of magnetic helicity, used to establish topological bounds for the magnetic field lines in ideal plasmas, and to understand the self-consistency of reconnected magnetic surfaces in non-axisymmetric configurations. After this theoretical discussion, we present some results on magnetohydrodynamic equilibrium and the use of analytical solutions to the Grad-Shafranov equation for describing real tokamak discharges based on the experimental diagnostics and realistic boundary conditions. We also compare the equilibrium reconstruction of a DIII-D discharge obtained with a numerical reconstruction routine, developed as part of this research, and the EFIT code used by several tokamak laboratories around the world. The magnetic topology and plasma profiles obtained with our method are in considerable agreement with the numerical reconstruction performed with the other code. Then, we introduce a simplified description of the generic non-axisymmetric magnetic field created by known sources and implement it numerically for describing the magnetic field due to external coils in tokamak devices. After that, we use this routines to develop a numerical procedure to adjust a suitable set of non-linear parameters of internal filamentary currents, which are intended to model the plasma response based on the magnetic field measurements outside the plasma. Finally, these methods are used to model the magnetic field created by a slowly rotating plasma instability in a real DIII-D discharge. The plasma response modeling is based on the magnetic probe measurements and allow us to calculate the magnetic field in arbitrary locations near the plasma edge. Using this information we determine the non-axisymmetric plasma edge through the magnetic invariant manifolds routine developed during this work. The intersection of the calculated invariant manifold with the tokamak chamber agrees considerably well with the heat flux measurements for the same discharge at the divertor plates, indicating the development of a rotating manifold due to the internal asymmetric plasma currents, giving quantitative support to our simplified description of the magnetic field and the plasma edge definition through the invariant manifolds. / Neste trabalho estuda-se a modelagem do campo magnético em configurações realistas de plasmas em equilíbrio não-axissimétrico e o fluxo de calor nos componentes em contato com o plasma em descargas de tokamaks com desviadores poloidais. Começa-se estabelecendo a relação entre configurações magnéticas arbitrárias e sistemas dinâmicos Hamiltonianos. Então aplicamos o conceito de helicidade magnética, que é usado para estabelecer limitações topológicas sobre as linhas de campo magnético em plasmas ideais, assim como para compreender a auto-consistência das superfícies magnéticas reconectadas em configurações não-axissimétricas. Após esta discussão teórica, apresentam-se alguns resultados sobre o equilíbrio magnetohidrodinâmico e o uso de soluções analíticas à equação de Grad-Shafranov para descrever descargas reais em tokamaks, com base em diagnósticos experimentais e condições de contorno realistas. Também realiza-se uma comparação entre a reconstrução do equilíbrio de uma descarga do DIII-D, obtida mediante uma rotina numérica desenvolvida para esta pesquisa, com a obtida mediante o código EFIT, usado amplamente em diversos tokamaks. Após isso, apresenta-se uma descrição simplificada do campo magnético não-axissimétrico, criado por fontes determinadas, e a sua implementação para descrever o campo magnético devido às correntes externas em tokamaks. Então, usam-se estas rotinas para desenvolver um procedimento numérico que ajusta um conjunto adequado de parâmetros não-lineares de correntes filamentares internas, com as quais pretende-se modelar a resposta do plasma com base nas medidas de campo magnético fora do plasma. Finalmente, estes métodos são utilizados para modelar o campo magnético criado por uma instabilidade com rotação lenta numa descarga do DIII-D. Com base nas medidas das sondas magnéticas é possível modelar os campos criados em regiões arbitrárias próximas da borda do plasma. Usando esta informação é possível determinar a borda não-axissimétrica do plasma mediante as invariantes magnéticas calculadas com a utilização de uma rotina desenvolvida durante este trabalho. A intersecção da superfície invariante com a câmara do tokamak coincide satisfatoriamente com as medidas de fluxo de calor nas placas do divertor para a mesma descarga, indicando o desenvolvimento de uma variedade giratória criada pelas correntes de plasma não-axissimétricas, e sustentando quantitativamente a nossa descrição simplificada do campo magnético, assim como a definição da borda do plasma mediante as invariantes magnéticas.

Dans les tokamaks, Sous l'application champ de électrique, les électrons sont accélérés et en même temps, ils subissent une force de friction due aux collisions avec les autres particules du plasma. Cependant, une fraction de la population totale d'électrons peuvent surmonter la force de friction et atteindre une vitesse proche de la vitesse lumière. Ces électrons relativistes sont découplés du plasma et sont appelés électrons runaway (ER). Ils peuvent apparaître lors des différentes phases d'une décharge de plasma. Par exemple, dans la phase de démarrage ou alors pendant les disruptions, au cours desquelles une fraction importante du courant plasma peut être convertie en ER ayant une énergie pouvant atteindre quelques dizaines de MeV. Les ER créés pendant la phase de perturbation peuvent causer des dommages aux premiers composants murs si un dépôt localisé de forte puissance se produit. ITER étant un tokamak de grande taille et un projet coûteux, la génération d'ER n'est pas souhaitable. La viabilité de la machine nécessite que les ER soient détectés en temps réel. La thèse fournit une étude détaillée dans cette direction pour le développement des deux principaux diagnostics sur ITER impliqués dans les mesures de paramètres pour les ER, à savoir, le moniteur de rayons X durs qui détecte le rayonnement de bremsstrahlung et les caméras visibles et infrarouges qui détectent le rayonnement synchrotron. Une solution de conception unique a été proposée pour le moniteur HXRM et est développée ici et optimisée. Pour les caméras, une modélisation des signaux est effectuée pour la première fois. Pour ce faire, un code de calcul a été développé et validé sur différents tokamaks. / In tokamaks, under the application of the electric field, a small fraction of the total electrons population can overcome collisional drag force and attain high velocity close to the speed of light. These relativistic electrons are called Runaway-Electrons (REs). The REs can occur during different phases of a plasma discharge. REs created during the disruptions phase can form a high energetic RE-beam that poses a risk to damage the first wall components if localized high power deposition takes place. ITER being a large size tokamak and an expensive project, generation of REs is not desirable during any phases of a plasma discharge. Detection of these REs and measurements of its parameters are important for the tokamak operation. Hence, RE diagnostics have to be in place to aid the commissioning of the disruption mitigation system and also for the post-event analysis to improve the reliability of RE avoidance. The present thesis gives a detailed study in this direction for the development of the two principal ITER Diagnostics involved in RE parameter measurements, namely the Hard X-Ray Monitor (HXRM) that detects bremsstrahlung radiation and the Visible and Infrared Cameras that detect synchrotron radiation. A unique design solution has been given for the HXRM and is developed, R&D tests were performed and optimized in line with this understanding. For the cameras, it is predicted for the first time which images and signal intensity can be expected. To achieve this, a simple but comprehensive code has been developed and validated on tokamaks that can predict RE parameters and corresponding diagnostic signals which may have further uses also in the context of RE avoidance.

L'objectif principal de la thèse est d'établir un modèle sous forme hamiltonienne à ports pour la dynamique du plasma dans les réacteurs de fusion de type tokamak, puis de démontrer le potentiel de cette approche pour aborder les problèmes d'intégration numérique et de commande non linéaire. Les bilans thermo-magnéto-hydrodynamiques, écrits sous forme hamiltonienne à ports à l'aide de structures Stokes-Dirac, conduisent à un modèle 3D “ multi-physique ” du plasma. Ensuite, un modèle 1D équivalent au modèle de diffusion résistive est obtenu en supposant les mêmes hypothèses d'équilibre quasi-statique et de symétries. Un schéma symplectique de réduction spatiale de ce modèle 1D qui préserve la structure du modèle et ses invariants est établi. Il ouvre la voie à des travaux ultérieurs de commande non linéaire fondés sur la structure géométrique d'interconnexion et les bilans du modèle. La commande IDA-PBC (Interconnection and Damping Assignment - Passivity Based Control) basée sur la passivité du modèle est d'abord synthétisée pour ce système en dimension finie. Finalement, une commande IDA-PBC associée avec la commande à la frontière est proposée pour le système en dimension infinie. Les controlleurs sont testés et validés avec les simulateurs des tokamak (METIS pour le Tore Supra de CEA/ Cadarache, et RAPTOR pour le TCV de l'EPFL Lausanne, Suisse). / The modelling and analysis of the plasma dynamics in tokamaks using the port-Hamiltonian approach is the main project purpose. Thermo-mMagnetohydrodynamics balances have been written in port-Hamiltonian form using Stokes-Dirac interconnection structures and 3D differential forms. A simplified 1D model for control has been derived using quasi-static and symmetry assumptions. It has been proved to be equivalent to a classical 1D control model: the resistive diffusion model for the poloidal magnetic flux. Then a geometric spatial integration scheme has been developped. It preserves both the symplecticity of the Dirac interconnection structure and physically conserved extensive quantities. This will allow coming works on energy-based approaches for the non linear control of the plasma dynamics.An Interconnection and Damping Assignment - Passivity Based Control (IDA-PBC) , the most general Port-Hamiltonian control, is chosen first to deal with the studied Tokamak system. It is based on a model made of the two coupled PDEs of resistive diffusion for the magnetic poloidal flux and of radial thermal diffusion. The used TMHD couplings are the Lorentz forces (with non-uniform resistivity) and the bootstrap current. The loop voltage at the plasma boundary, the total external current and the plasma heating power are considered as controller outputs. Due to the actuator constraints which imply to have a physically feasible current profile deposits, a feedforward control is used to ensure the compatibility with the actuator physical capability. Then, the IDA-PBC controllers, both finite-dimensional and infinite-dimensional, are designed to improve the system stabilization and convergence speed. The proposed works are validated against the simulation data obtained from the Tore-Supra WEST (CEA/Cadarache, France) test case and from RAPTOR code for the TCV real-time control system (CRPP/ EPFL, Lausanne, Switzerland).

In this work we study the magnetic field modeling of realistic non-axisymmetric plasma equilibrium configurations and the heat flux patterns on the plasma facing components of tokamak divertor discharges. We start by establishing the relation between generic magnetic configurations and Hamiltonian dynamical systems. We apply the concept of magnetic helicity, used to establish topological bounds for the magnetic field lines in ideal plasmas, and to understand the self-consistency of reconnected magnetic surfaces in non-axisymmetric configurations. After this theoretical discussion, we present some results on magnetohydrodynamic equilibrium and the use of analytical solutions to the Grad-Shafranov equation for describing real tokamak discharges based on the experimental diagnostics and realistic boundary conditions. We also compare the equilibrium reconstruction of a DIII-D discharge obtained with a numerical reconstruction routine, developed as part of this research, and the EFIT code used by several tokamak laboratories around the world. The magnetic topology and plasma profiles obtained with our method are in considerable agreement with the numerical reconstruction performed with the other code. Then, we introduce a simplified description of the generic non-axisymmetric magnetic field created by known sources and implement it numerically for describing the magnetic field due to external coils in tokamak devices. After that, we use this routines to develop a numerical procedure to adjust a suitable set of non-linear parameters of internal filamentary currents, which are intended to model the plasma response based on the magnetic field measurements outside the plasma. Finally, these methods are used to model the magnetic field created by a slowly rotating plasma instability in a real DIII-D discharge. The plasma response modeling is based on the magnetic probe measurements and allow us to calculate the magnetic field in arbitrary locations near the plasma edge. Using this information we determine the non-axisymmetric plasma edge through the magnetic invariant manifolds routine developed during this work. The intersection of the calculated invariant manifold with the tokamak chamber agrees considerably well with the heat flux measurements for the same discharge at the divertor plates, indicating the development of a rotating manifold due to the internal asymmetric plasma currents, giving quantitative support to our simplified description of the magnetic field and the plasma edge definition through the invariant manifolds. / Neste trabalho estuda-se a modelagem do campo magnético em configurações realistas de plasmas em equilíbrio não-axissimétrico e o fluxo de calor nos componentes em contato com o plasma em descargas de tokamaks com desviadores poloidais. Começa-se estabelecendo a relação entre configurações magnéticas arbitrárias e sistemas dinâmicos Hamiltonianos. Então aplicamos o conceito de helicidade magnética, que é usado para estabelecer limitações topológicas sobre as linhas de campo magnético em plasmas ideais, assim como para compreender a auto-consistência das superfícies magnéticas reconectadas em configurações não-axissimétricas. Após esta discussão teórica, apresentam-se alguns resultados sobre o equilíbrio magnetohidrodinâmico e o uso de soluções analíticas à equação de Grad-Shafranov para descrever descargas reais em tokamaks, com base em diagnósticos experimentais e condições de contorno realistas. Também realiza-se uma comparação entre a reconstrução do equilíbrio de uma descarga do DIII-D, obtida mediante uma rotina numérica desenvolvida para esta pesquisa, com a obtida mediante o código EFIT, usado amplamente em diversos tokamaks. Após isso, apresenta-se uma descrição simplificada do campo magnético não-axissimétrico, criado por fontes determinadas, e a sua implementação para descrever o campo magnético devido às correntes externas em tokamaks. Então, usam-se estas rotinas para desenvolver um procedimento numérico que ajusta um conjunto adequado de parâmetros não-lineares de correntes filamentares internas, com as quais pretende-se modelar a resposta do plasma com base nas medidas de campo magnético fora do plasma. Finalmente, estes métodos são utilizados para modelar o campo magnético criado por uma instabilidade com rotação lenta numa descarga do DIII-D. Com base nas medidas das sondas magnéticas é possível modelar os campos criados em regiões arbitrárias próximas da borda do plasma. Usando esta informação é possível determinar a borda não-axissimétrica do plasma mediante as invariantes magnéticas calculadas com a utilização de uma rotina desenvolvida durante este trabalho. A intersecção da superfície invariante com a câmara do tokamak coincide satisfatoriamente com as medidas de fluxo de calor nas placas do divertor para a mesma descarga, indicando o desenvolvimento de uma variedade giratória criada pelas correntes de plasma não-axissimétricas, e sustentando quantitativamente a nossa descrição simplificada do campo magnético, assim como a definição da borda do plasma mediante as invariantes magnéticas.

Orientador: Ricardo Egydio de Carvalho / Banca: Iberê Luiz Caldas / Banca: Marisa Roberto / Resumo: Investigamos nesta dissertação a introdução de uma barreira dinâmica em diferentes sistemas físicos caóticos, a fim de analisar a influência que esta barreira causa na dinâmica e topologia destes sistemas. A barreira principal deste estudo é a barreira denominada Toro Robusto, que nada mais é do que uma curva invariante no espaço de fases em meio a estruturas de ressonância, mares de caos, etc. A barreira Toro Robusto bloqueia a difusão caótica no espaço de fases associado ao sistema físico, e causa também uma estabilização em sua vizinhança linear. Introduziremos Toros Robustos em vários tipos de sistemas dinâmicos, como por exemplo, em uma Hamiltoniana "Toy Model" a fim de entender o seu efeito no processo de reconexão ou "overlap" de ressonâncias isócronas. Toros Robustos quebrando a dimerização de cadeias de ressonância também foram estudados no mapa padrão "não-twist". O bloqueio da difusão de Arnold no mapa padrão acoplado também foi mostrado, assim como, a introdução de Toros Robustos em sistemas utilizados em física de plasmas, como meio de controle de caos em plasma confinado em Tokamak. Outra barreira apresentada aqui é a barreira do tipo "meander" que surge através do processo de reconexão de ressonâncias no espaço de fases. Introduziremos um novo mapa discreto que chamamos de Mapa padrão "não-twist" labiríntico, que apresenta múltiplas regiões de barreiras "meanders" por todo o espaço de fases / Abstract: We investigated in this work the introduction of a dynamical barrier in different chaotic physical systems in order to analyze the influence that it causes in the topology and in the dynamics of them. The main barrier studied here is called Robust Tori which is an invariant curve in the phase space permeated by resonance structures and chaotic seas. The Robust Torus barrier blocks the chaotic diffusion in the phase space of the associated physical system, and it also causes a linear stabilization in its neighborhood. Robust Tori will be introduced in several types of dynamic systems, such as in a Toy Model Hamiltonian in order to understand their effect on the reconnection process or overlap of isochronous resonances. The breakdown of resonance dimerization by Robust Tori was also studied using the nontwist standard map. The blocking of Arnold diffusion in the coupled standard map was also shown, as well as the introduction of Robust Tori in relevant models for plasma physics as a tool for controlling chaos in confined plasmas in Tokamaks. Another barrier, which is presented here, is the meander barrier that emerges through the reconnection process of resonances in phase space. We will also introduce a new discrete map, which we call labyrinthic standard non-twist map that shows multiple regions of meanders barriers around the phase space / Mestre

L'injection d'ondes radiofréquence dans un plasma de tokamak afin d'y générer le courant toroïdal répond à une double exigence. Premièrement, la nature non inductive de la méthode évite le recours aux courants variables circulant dans les bobines, peu compatibles avec l'opération stationnaire d'un futur réacteur. Par ailleurs, il est reconnu que la principale limitation des performances d'un plasma de fusion est causée par la turbulence électromagnétique. Celle-ci peut toutefois être réduite, voire supprimée, en optimisant le profil de courant, ce qu'autorise précisément l'emploi des ondes, dans le cadre des scénarios avancés. Cette thèse traite de l'utilisation de l'onde cyclotronique électronique (EC) en vue de contrôler le profil de courant. S'agissant d'une question cruciale conditionnant l'usage de cette onde dans les plasma chauds, l'effet de la température finie sur la polarisation de l'onde est d'abord étudié dans divers régimes. D'autre part, dans les scénarios avancés, l'association des ondes EC et hybride basse (LH) est prometteuse, du fait de leurs caractéristiques complémentaires. Une large partie de ce travail est donc consacrée à l'étude théorique, numérique et expérimentale des décharges combinées. Les résultats obtenus, parmi lesquels la démonstration analytique d'un effet de synergie entre les deux ondes, montrent clairement l'intérêt de ces scénarios et motivent la mise au point de nouvelles expériences.

Resonant Magnetic Perturbation (RMP) fields produced by external control coils are considered a viable option for the suppression of Edge Localized Modes (ELMs) in present and future tokamaks. Repeated reversals of the toroidal phase of the I-coil magnetic field in RMP shot 147170 on DIII-D has generated uniquely different edge pedestal profiles, implying different edge transport phenomena. The causes, trends, and implications of RMP toroidal phase reversal on edge transport is analyzed by comparing various parameters at 0 and 60 degree toroidal phases, with an I-coil mode number of n=3. An analysis of diffusive and non-diffusive transport effects of these magnetic perturbations it the plasma edge pedestal for this RMP shot is characterized by interpreting the ion and electron heat diffusivities, angular momentum transport frequencies, ion diffusion coefficients, and pinch velocities for both phases.

En el estudio de la fusión por confinamiento magnético es crucial conocer con detalle las condiciones físicas del plasma confinado. En este sentido los códigos de simulación son una herramienta muy útil para interpretar las propiedades del plasma. A lo largo de esta tesis se ha estudiado las propiedades del plasma de tokamaks y stellarators y se ha contribuido al desarrollo del código de transporte PRETOR, creado en el JET. Estos estudios se han realizado dentro de la colaboración existente entre el DFEN de la UPC y el Laboratorio Nacional de Fusión del CIEMAT.El código PRETOR permite simular la variación radial y la evolución temporal de las principales magnitudes físicas de un plasma de fusión. Para valorar la bondad de los modelos implementados en el código y determinar la corrección de sus resultados los resultados de la simulación de las magnitudes más relevantes de diversas descargas se comparan con los datos experimentales. De esta comparación se deduce que las magnitudes simuladas presentan un buen acuerdo con los datos experimentales aunque debe destacarse que los errores experimentales son bastante grandes y no siempre están disponibles.En el desarrollo de esta tesis también se ha aplicado el código PRETOR para el estudio de la parada de emergencia de ITER-DDR, con 1500 MW de potencia de fusión funcionando en un estado estacionario en ignición. Se ha estudiado con detalle la evolución del plasma durante una parada implementada mediante la interrupción del suministro de combustible. Este método se muestra efectivo para reducir la potencia de fusión en unas decenas de segundos evitando una disrupción del plasma en todo el escenario.Se han realizado diversas modificaciones en el código para hacerlo más versátil y adaptarlo a las necesidades específicas que han ido surgiendo. La modificación más destacada es la incorporación a PRETOR de una rutina más elaborada que la inicialmente implementada en el código para el cálculo de la deposición de potencia adicional por inyección de haces neutros.En los últimos años la investigación de stellarators ha adquirido una creciente importancia debido a las ventajas que presentan este tipo de dispositivos. Además hay que destacar que el programa de fusión por confinamiento magnético español se centra en el Heliac Flexible TJ-II de la asociación EURATOM-CIEMAT. Por estos motivos y para disponer de un código que sea capaz de simular descargas de stellarators, PRETOR ha sido modificado de forma substancial para crear la versión PRETOR-Stellarator. Esta modificación del código presenta un aspecto inédito de la tesis en el ámbito internacional.Los módulos que han sido adaptados afectan a la configuración magnética, al transporte de energía y partículas y a la geometría del plasma. Una vez modificado el código, éste se ha aplicado al análisis del plasma de Heliac Flexible TJ-II. El objetivo de este estudio es comprobar que las modificaciones introducidas en PRETOR permiten simular correctamente descargas de stellarators. Los resultados obtenidos con PRETOR se comparan con los obtenidos con el código PROCTR, un código utilizado en el CIEMAT para el estudio del TJ-II y de otros stellarators.Los resultados demuestran que los perfiles de temperatura se reproducen correctamente, estos resultados son similares a los obtenidos con PROCTR. Los valores experimentales de la potencia radiada y de la energía almacenada en el plasma han sido satisfactoriamente predichos mediante la simulación de PRETOR-Stellarator.El camino abierto por esta tesis permitirá ahondar en el estudio del plasma del heliac flexible TJ-II, el código está disponible para utilizarse en este dispositivo e incorporar nuevas modificaciones para adaptarlo a las necesidades que vayan apareciendo. / In the study of the fusion by magnetic confinement it is crucial to know the physical behaviour of the confined plasma with detail. In this sense the simulation codes are a very useful tool to analyse the properties of the plasma. Throughout this thesis the properties of the plasma of tokamaks and stellarators has been studied and the transport code PRETOR, created in the JET, has been developed. These studies have been developed within the existing collaboration between the DFEN in the UPC and the Laboratorio Nacional de Fusión of the CIEMAT.The PRETOR code allows the simulation of the radial variation and the temporal evolution of the main physical magnitudes of a fusion plasma. The results of the simulation of the most important magnitudes of diverse shots are compared with the experimental data in order to evaluate the goodness of the models implemented in the code and to determine the correction of their results. From this comparison it's deduced that the simulated magnitudes present a good agreement with the experimental data although it must be stand out that the experimental errors are quite large and they not always are available.In the development of this thesis the code PRETOR has also been applied to the study of the shutdown emergency of ITER-DDR, with 1500 MW of fusion power operating in an ignition stationary state. The evolution of the plasma has studied with detail during the shutdown implemented by means of the interruption of the fuel injection. This method is effective to reduce the fusion power in tens of seconds avoiding a plasma disrupción in all the scenarios. Several modifications have been made in the code in order to make it more versatile and to adapt it to the specific arisen necessities. The most outstanding modification is the incorporation to PRETOR of more complex routine than that initially implemented in the code for the calculation of the additional power deposition by neutral beam injection.In the last years the investigation of stellarators has acquired an increasing importance due to the advantages that this type of devices presents. Besides it is necessary to emphasise that the Spanish program of fusion by magnetic confinement is focused in Flexible Heliac TJ-II in the asociación EURATOM-CIEMAT. For these reasons and in order to have a code able to simulate stellarators shots, PRETOR has been modified of substantial form to create the PRETOR-Stellarator version. This modification of the code presents an new aspect of the thesis in the international scope.The modules that have been adapted affect to the magnetic configuration, the transport of energy and particles and to the plasma geometry. Once the code has been modified, it has been applied to the analysis of TJ-II Flexible Heliac plasmas. The objective of this study is to verify that the modifications introduced in PRETOR allow to simulate correctly stellarators discharges. The results obtained with PRETOR are compared with that ones obtained with PROCTR code, a code used in the CIEMAT for the study of the TJ-II and others stellarators.The results demonstrate that the temperature profiles are reproduced correctly, these results are similar to the obtained ones with PROCTR. The experimental values of the radiated power and the stored energy in the plasma have been predicted satisfactorily by means of the simulation with PRETOR-Stellarator.The way open by this thesis will allow studying in depth the plasma of TJ-II flexible heliac, the code is available to be used in this device and to incorporate new modifications to adapt it to the necessities that will appear.