Tokamak magnetic field lines described by simple maps

The magnetic field line structure in a tokamak can be obtained by direct numerical integration of the field line equations. However, this is a lengthy procedure and the analysis of the solution may be very time-consuming. Otherwise we can use simple two-dimensional, area-preserving maps, obtained either by approximations of the magnetic field line equations, or from dynamical considerations. These maps can be quickly iterated, furnishing solutions that mirror the ones obtained from direct numerical integration, and which are useful when long-term studies of field line behavior are necessary (e.g. in diffusion calculations). In this work we focus on a set of simple tokamak maps for which these advantages are specially pronounced.

Self-consistent equilibrium calculation through a direct variational technique in tokamak plasmas

A self-consistent equilibrium calculation, valid for arbitrary aspect ratio tokamaks, is obtained through a direct variational technique that reduces the equilibrium solution, in general obtained from the 2D Grad-Shafranov equation, to a 1D problem in the radial flux coordinate rho. The plasma current profile is supposed to have contributions of the diamagnetic, Pfirsch-Schluter and the neoclassical ohmic and bootstrap currents. An iterative procedure is introduced into our code until the flux surface averaged toroidal current density (J(T)), converges to within a specified tolerance for a given pressure profile and prescribed boundary conditions. The convergence criterion is applied between the (J(T)) profile used to calculate the equilibrium through the variational procedure and the one that results from the equilibrium and given by the sum of all current components. The ohmic contribution is calculated from the neoclassical conductivity and from the self-consistently determined loop voltage in order to give the prescribed value of the total plasma current. The bootstrap current is estimated through the full matrix Hirshman-Sigmar model with the viscosity coefficients as proposed by Shaing, which are valid in all plasma collisionality regimes and arbitrary aspect ratios. The results of the self-consistent calculation are presented for the low aspect ratio tokamak Experimento Tokamak Esferico. A comparison among different models for the bootstrap current estimate is also performed and their possible Limitations to the self-consistent calculation is analysed.

Tokamak TCABR: Acquisition system, data analysis, and remote participation using MDSplus

Each plasma physics laboratory has a proprietary scheme to control and data acquisition system. Usually, it is different from one laboratory to another. It means that each laboratory has its own way to control the experiment and retrieving data from the database. Fusion research relies to a great extent on international collaboration and this private system makes it difficult to follow the work remotely. The TCABR data analysis and acquisition system has been upgraded to support a joint research programme using remote participation technologies. The choice of MDSplus (Model Driven System plus) is proved by the fact that it is widely utilized, and the scientists from different institutions may use the same system in different experiments in different tokamaks without the need to know how each system treats its acquisition system and data analysis. Another important point is the fact that the MDSplus has a library system that allows communication between different types of language (JAVA, Fortran, C, C++, Python) and programs such as MATLAB, IDL, OCTAVE. In the case of tokamak TCABR interfaces (object of this paper) between the system already in use and MDSplus were developed, instead of using the MDSplus at all stages, from the control, and data acquisition to the data analysis. This was done in the way to preserve a complex system already in operation and otherwise it would take a long time to migrate. This implementation also allows add new components using the MDSplus fully at all stages. (c) 2012 Elsevier B.V. All rights reserved.

A new generation of real-time systems in the JET tokamak

Recently a new recipe for developing and deploying real-time systems has become increasingly adopted in the JET tokamak. Powered by the advent of x86 multi-core technology and the reliability of the JET’s well established Real-Time Data Network (RTDN) to handle all real-time I/O, an official Linux vanilla kernel has been demonstrated to be able to provide realtime performance to user-space applications that are required to meet stringent timing constraints. In particular, a careful rearrangement of the Interrupt ReQuests’ (IRQs) affinities together with the kernel’s CPU isolation mechanism allows to obtain either soft or hard real-time behavior depending on the synchronization mechanism adopted. Finally, the Multithreaded Application Real-Time executor (MARTe) framework is used for building applications particularly optimised for exploring multicore architectures. In the past year, four new systems based on this philosophy have been installed and are now part of the JET’s routine operation. The focus of the present work is on the configuration and interconnection of the ingredients that enable these new systems’ real-time capability and on the impact that JET’s distributed real-time architecture has on system engineering requirements, such as algorithm testing and plant commissioning. Details are given about the common real-time configuration and development path of these systems, followed by a brief description of each system together with results regarding their real-time performance. A cycle time jitter analysis of a user-space MARTe based application synchronising over a network is also presented. The goal is to compare its deterministic performance while running on a vanilla and on a Messaging Real time Grid (MRG) Linux kernel.

Estudo de efeitos da polarização eletrostática periférica no Tokamak TCABR

Efeitos da polarização eletrostática de eletrodos na periferia de tokamaks têm sido investigados em pequenos tokamaks e mesmo em alguns tokamaks de grande porte. Em geral as experiências são realizadas em condições em que bifurcação do campo elétrico radial é obtida, processo este identificado como modo H de polarização. No Tokamak TCABR, as experiências indicam que o confinamento aumenta para tensões aplicadas até +300 volts, atingindo um máximo de duas vezes o tempo de confinamento do modo L, mas sem bifurcação. Indícios de bifurcação foram notados com +400 V de polarização, mas a descarga termina devido à excitação da atividade MHD, ainda sob investigação. No presente trabalho, a pesquisa é aprofundada com a utilização de uma sonda de Langmuir com 18 pinos dispostos em duas fileiras sob a forma de um ancinho (rake probe) o que permite a medição da temperatura, densidade e flutuação de potencial ao longo do raio menor na periferia do Tokamak. A resolução temporal desse sistema é de cerca de 0,5 ms, para a temperatura, e 5 microssegundos para densidade e potencial flutuante do plasma. Outra sonda eletrostática com 5-pinos na mesma posição radial, mas em diferentes posições poloidal e toroidal foi usada para medições de turbulência e transporte de partículas. Os efeitos da polarização foram investigados e indicam que os níveis de turbulência e transporte começam a diminuir entre +150 e +200 V e para +300 V chegam a atingir uma quase supressão. Nesse mesmo intervalo de tensão a densidade começa a aumentar e para +300 V chega a ser um fator de aproximadamente 2. Quanto ao perfil de temperatura a variação é pouco significativa, mas as incertezas das medidas são maiores. Esses dados são compatíveis com a criação de uma barreira de transporte na região entre o eletrodo em r = 17 cm e o limitador em a = 18 cm. Além disso, o campo elétrico radial mostra forte cisalhamento nessa região. Tomando o início da subida do potencial flutuante como origem de uma escala de tempo, o atraso temporal do início da subida da densidade de elétrons e o atraso do início do decréscimo do transporte de partículas foram medidos. Os resultados são 50 microssegundos para a densidade de elétrons e 60 microssegundos para o transporte de partículas. A questão dos limiares de potência é discutida no texto. Os dados desta experiência indicam que o campo elétrico radial desempenha o papel principal para a melhoria do confinamento.

Instabilidades MHD no Tokamak TCABR

Este trabalho descreve o estudo das instabilidades magneto-hidro-dinâmicas (MHD) comumente observadas nas descargas elétricas de plasma no tokamak TCABR, do Instituto de Física da USP. Dois diagnósticos principais foram empregados para observar essas instabilidades: um conjunto poloidal de 24 bobinas magnéticas (bobinas de Mirnov) colocadas próximas à borda do plasma e um medidor de emissões na faixa do Ultra Violeta e de raios X moles com 20 canais (sistema SXR), cujo circuito de condicionamento de sinais foi aprimorado como parte deste trabalho. Esses diagnósticos foram escolhidos porque fornecem informações complementares, uma vez que o sistema SXR observa a parte central da coluna de plasma, enquanto as bobinas de Mirnov detectam as instabilidades MHD na região mais externa da coluna. As informações coletadas por esses diagnósticos foram submetidas à análise espectral com resolução temporal e espacial, possibilitando determinar a evolução das características espectrais e espaciais das instabilidades MHD observadas. Essas análises revelaram que durante a etapa inicial da formação do plasma (quando a corrente de plasma ainda está aumentando) ilhas magnéticas com números de onda decrescente, identificadas como sendo modos kink de borda, são detectadas nas bobinas de Mirnov. Após a formação do plasma, quando os parâmetros de equilíbrio estão relativamente estáveis (platô), oscilações são detectadas tanto nas bobinas de Mirnov quanto no sistema de SXR, indicando a presença de instabilidades MHD em toda a coluna de plasma. Em geral as oscilações medidas nas bobinas de Mirnov tem baixa amplitude e correspondem a pequenas ilhas magnéticas que foram identificadas como sendo modos de ruptura (modos tearing). Por outro lado, as instabilidades na região central foram identificadas como dentes de serra, que correspondem a relaxações periódicas da região interna à superfície magnética com fator de segurança q=1 e que são acompanhadas de oscilações precursoras, cuja amplitude depende da fase do ciclo de relaxação. Devido à essa modulação de amplitude, aparecem picos de frequência satélite nos espectrogramas dos sinais do SXR. Além disso, devido ao fato dos ciclos de relaxação não serem sinusoidais, os harmônicos da frequência de relaxação também aparecem nesses espectrogramas. No entanto, em muitas descargas do TCABR, a intensidade das oscilações medidas nas bobinas de Mirnov aumentam significativamente durante o platô, com efeitos sobre a frequência de todas as instabilidades MHD, até mesmo sobre os dentes de serra localizados na região central da coluna. Em todos os casos, observou-se que durante o platô a frequência das ilhas magnéticas coincide com a frequência das oscilações precursoras do dente de serra, apesar de serem duas instabilidades distintas, localizadas em posições radiais muito diferentes. Essa coincidência de frequências possibilitou descrever a evolução em frequência de todas as oscilações detectadas em diversos diagnósticos com base em apenas duas frequências básicas: a dos ciclos de relaxação dente de serra e a das ilhas magnéticas.

Physical characterization and numerical modeling of a WEST-like tokamak divertor plasma

Il presente elaborato è incentrato sulla modellizzazione del plasma di bordo nei dispositivi per la produzione di energia da fusione nucleare noti come tokamak. La tecnologia che nel corso di tutta la seconda metà del XX secolo fino ad oggi è stata sviluppata a questo fine deve necessariamente scontrarsi con alcuni limiti. Nei tokamak il confinamento del plasma è di tipo magnetico e vincola le particelle a muoversi di moto elicoidale all'interno del vessel, tuttavia il confinamento non risulta perfetto e parte dell'energia si scarica sulle pareti della camera, rischiando pertanto di fondere i materiali. Alcune strategie possono essere messe in atto per limitare questo problema, per esempio agendo sulla geometria del tokamak, oppure sulla fisica, inducendo nel plasma una data concentrazione di impurezze che ionizzino irraggiando parte dell'energia di plasma. Proprio tale meccanismo di perdita è stato simulato in un modello monodimensionale di plasma monofluido di bordo. I risultati del codice numerico relativo al modello dimostrano che per concentrazioni di impurezze crescenti è possibile diminuire in modo significativo flusso di calore e temperatura al divertore. Per di più risulta possibile controllare la posizione del fronte di irraggiamento per mezzo di parametri di controllo del plasma quali la pressione. Si osserva inoltre l'insorgere del cosiddetto fenomeno di biforcazione alle basse temperature di divertore, fenomeno in cui il plasma si comporta in modo instabile a causa di fenomeni fisici tipici delle basse energie ("detachment") e a seguito del quale può improvvisamente spegnersi (disruzione). Infine lo stesso modello è stato migliorato inserendo l'ipotesi di plasma bifluido. Anche per gli ioni viene osservato il fenomeno di biforcazione. I risultati numerici evidenziano le dinamiche dello scambio energetico fra le specie gettando le basi di una progettazione efficiente della chimica del plasma finalizzata al raffreddamento del divertore.

Description of the ITER Tokamak Cooling Water System

In questa tesi ho inizialmente esposto cenni teorici sulle reazioni di fusione nucleare e le motivazioni che hanno spinto la comunità scientifica verso la ricerca di questa nuova fonte energetica. Ho descritto il progetto ITER nei suoi obiettivi e nei principi di funzionamento di un reattore di tipo Tokamak e di tutti i componenti principali dell'intero impianto. In primo piano, mi sono focalizzato sul sistema di raffreddamento primario ad acqua del Tokamak (TCWS), con una prima panoramica sui suoi sottosistemi descrivendo i loro obiettivi, quali asportazione di calore e sicurezza dell'impianto. Successivamente ho analizzato nello specifico i particolari tecnici dei principali sottosistemi quali i vari circuiti di asportazione primaria del calore (PHTS Loops) dei diversi componenti del Tokamak, il Vacuum Vessel, il First Wall Blanket, il Divertor e il Neutral Beam Injector; ho esaminato i processi di controllo della qualità e del volume del fluido refrigerante nei circuiti (CVCS); ed infine le funzioni e le caratteristiche dei sistemi di drenaggio e di riempimento dei circuiti con i propri serbatoi ordinari e di sicurezza, e del sistema di asciugatura del fluido refrigerante con le sue diverse modalità operative.

Tokamak experimental power reactor conceptual design /

"August, 1976."

Feasibility study of a fission-suppressed Tokamak fusion breeder /

"UCID-20154"--Cover.

Tokamak experimental power reactor studies /

"Prepared for the U.S. Energy Research and Development Administration under Contract W-31-109-Eng-38"--Cover.

Real-time control of a Tokamak plasma using neural networks

This paper presents results from the first use of neural networks for the real-time feedback control of high temperature plasmas in a Tokamak fusion experiment. The Tokamak is currently the principal experimental device for research into the magnetic confinement approach to controlled fusion. In the Tokamak, hydrogen plasmas, at temperatures of up to 100 Million K, are confined by strong magnetic fields. Accurate control of the position and shape of the plasma boundary requires real-time feedback control of the magnetic field structure on a time-scale of a few tens of microseconds. Software simulations have demonstrated that a neural network approach can give significantly better performance than the linear technique currently used on most Tokamak experiments. The practical application of the neural network approach requires high-speed hardware, for which a fully parallel implementation of the multi-layer perceptron, using a hybrid of digital and analogue technology, has been developed.

Interatomic potentials for fusion reactor material simulations

Fusion energy is a clean and safe solution for the intricate question of how to produce non-polluting and sustainable energy for the constantly growing population. The fusion process does not result in any harmful waste or green-house gases, since small amounts of helium is the only bi-product that is produced when using the hydrogen isotopes deuterium and tritium as fuel. Moreover, deuterium is abundant in seawater and tritium can be bred from lithium, a common metal in the Earth's crust, rendering the fuel reservoirs practically bottomless. Due to its enormous mass, the Sun has been able to utilize fusion as its main energy source ever since it was born. But here on Earth, we must find other means to achieve the same. Inertial fusion involving powerful lasers and thermonuclear fusion employing extreme temperatures are examples of successful methods. However, these have yet to produce more energy than they consume. In thermonuclear fusion, the fuel is held inside a tokamak, which is a doughnut-shaped chamber with strong magnets wrapped around it. Once the fuel is heated up, it is controlled with the help of these magnets, since the required temperatures (over 100 million degrees C) will separate the electrons from the nuclei, forming a plasma. Once the fusion reactions occur, excess binding energy is released as energetic neutrons, which are absorbed in water in order to produce steam that runs turbines. Keeping the power losses from the plasma low, thus allowing for a high number of reactions, is a challenge. Another challenge is related to the reactor materials, since the confinement of the plasma particles is not perfect, resulting in particle bombardment of the reactor walls and structures. Material erosion and activation as well as plasma contamination are expected. Adding to this, the high energy neutrons will cause radiation damage in the materials, causing, for instance, swelling and embrittlement. In this thesis, the behaviour of a material situated in a fusion reactor was studied using molecular dynamics simulations. Simulations of processes in the next generation fusion reactor ITER include the reactor materials beryllium, carbon and tungsten as well as the plasma hydrogen isotopes. This means that interaction models, {\it i.e. interatomic potentials}, for this complicated quaternary system are needed. The task of finding such potentials is nonetheless nearly at its end, since models for the beryllium-carbon-hydrogen interactions were constructed in this thesis and as a continuation of that work, a beryllium-tungsten model is under development. These potentials are combinable with the earlier tungsten-carbon-hydrogen ones. The potentials were used to explain the chemical sputtering of beryllium due to deuterium plasma exposure. During experiments, a large fraction of the sputtered beryllium atoms were observed to be released as BeD molecules, and the simulations identified the swift chemical sputtering mechanism, previously not believed to be important in metals, as the underlying mechanism. Radiation damage in the reactor structural materials vanadium, iron and iron chromium, as well as in the wall material tungsten and the mixed alloy tungsten carbide, was also studied in this thesis. Interatomic potentials for vanadium, tungsten and iron were modified to be better suited for simulating collision cascades that are formed during particle irradiation, and the potential features affecting the resulting primary damage were identified. Including the often neglected electronic effects in the simulations was also shown to have an impact on the damage. With proper tuning of the electron-phonon interaction strength, experimentally measured quantities related to ion-beam mixing in iron could be reproduced. The damage in tungsten carbide alloys showed elemental asymmetry, as the major part of the damage consisted of carbon defects. On the other hand, modelling the damage in the iron chromium alloy, essentially representing steel, showed that small additions of chromium do not noticeably affect the primary damage in iron. Since a complete assessment of the response of a material in a future full-scale fusion reactor is not achievable using only experimental techniques, molecular dynamics simulations are of vital help. This thesis has not only provided insight into complicated reactor processes and improved current methods, but also offered tools for further simulations. It is therefore an important step towards making fusion energy more than a future goal.

Multiscale study on hydrogen mobility in metallic fusion divertor material

For achieving efficient fusion energy production, the plasma-facing wall materials of the fusion reactor should ensure long time operation. In the next step fusion device, ITER, the first wall region facing the highest heat and particle load, i.e. the divertor area, will mainly consist of tiles based on tungsten. During the reactor operation, the tungsten material is slowly but inevitably saturated with tritium. Tritium is the relatively short-lived hydrogen isotope used in the fusion reaction. The amount of tritium retained in the wall materials should be minimized and its recycling back to the plasma must be unrestrained, otherwise it cannot be used for fueling the plasma. A very expensive and thus economically not viable solution is to replace the first walls quite often. A better solution is to heat the walls to temperatures where tritium is released. Unfortunately, the exact mechanisms of hydrogen release in tungsten are not known. In this thesis both experimental and computational methods have been used for studying the release and retention of hydrogen in tungsten. The experimental work consists of hydrogen implantations into pure polycrystalline tungsten, the determination of the hydrogen concentrations using ion beam analyses (IBA) and monitoring the out-diffused hydrogen gas with thermodesorption spectrometry (TDS) as the tungsten samples are heated at elevated temperatures. Combining IBA methods with TDS, the retained amount of hydrogen is obtained as well as the temperatures needed for the hydrogen release. With computational methods the hydrogen-defect interactions and implantation-induced irradiation damage can be examined at the atomic level. The method of multiscale modelling combines the results obtained from computational methodologies applicable at different length and time scales. Electron density functional theory calculations were used for determining the energetics of the elementary processes of hydrogen in tungsten, such as diffusivity and trapping to vacancies and surfaces. Results from the energetics of pure tungsten defects were used in the development of an classical bond-order potential for describing the tungsten defects to be used in molecular dynamics simulations. The developed potential was utilized in determination of the defect clustering and annihilation properties. These results were further employed in binary collision and rate theory calculations to determine the evolution of large defect clusters that trap hydrogen in the course of implantation. The computational results for the defect and trapped hydrogen concentrations were successfully compared with the experimental results. With the aforedescribed multiscale analysis the experimental results within this thesis and found in the literature were explained both quantitatively and qualitatively.

Studies of cryocooler based cryosorption pump with activated carbon panels operating at 11K

Cryosorption pump is the only solution for pumping helium and hydrogen in fusion reactors. It is chosen because it offers highest pumping speed as well as the only suitable pump for the harsh environments in a tokamak. Towards the development of such cryosorption pumps, the optimal choice of the right activated carbon panels is essential. In order to characterize the performance of the panels with indigenously developed activated carbon, a cryocooler based cryosorption pump with scaled down sizes of panels is experimented. The results are compared with the commercial cryopanel used in a CTI cryosorption (model: Cryotorr 7) pump. The cryopanel is mounted on the cold head of the second stage GM cryocooler which cools the cryopanel down to 11K with first stage reaching about similar to 50K. With no heat load, cryopump gives the ultimate vacuum of 2.1E-7 mbar. The pumping speed of different gases such as nitrogen, argon, hydrogen, helium are tested both on indigenous and commercial cryopanel. These studies serve as a bench mark towards the development of better cryopanels to be cooled by liquid helium for use with tokamak.