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.

Integrable maps with non-trivial topology: application to divertor configurations

We explore a method for constructing two-dimensional area-preserving, integrable maps associated with Hamiltonian systems, with a given set of fixed points and given invariant curves. The method is used to find an integrable Poincare map for the field lines in a large aspect ratio tokamak with a poloidal single-null divertor. The divertor field is a superposition of a magnetohydrodynamic equilibrium with an arbitrarily chosen safety factor profile, with a wire carrying an electric current to create an X-point. This integrable map is perturbed by an impulsive perturbation that describes non-axisymmetric magnetic resonances at the plasma edge. The non-integrable perturbed map is applied to study the structure of the open field lines in the scrape-off layer, reproducing the main transport features obtained by integrating numerically the magnetic field line equations, such as the connection lengths and magnetic footprints on the divertor plate.

Divertor map with freedom of geometry and safety factor profile

An explicit, area-preserving and integrable magnetic field line map for a single-null divertor tokamak is obtained using a trajectory integration method to represent equilibrium magnetic surfaces. The magnetic surfaces obtained from the map are capable of fitting different geometries with freely specified position of the X-point, by varying free model parameters. The safety factor profile of the map is independent of the geometric parameters and can also be chosen arbitrarily. The divertor integrable map is composed of a nonintegrable map that simulates the effect of external symmetry-breaking resonances, so as to generate a chaotic region near the separatrix passing through the X-point. The composed field line map is used to analyze escape patterns (the connection length distribution and magnetic footprints on the divertor plate) for two equilibrium configurations with different magnetic shear profiles at the plasma edge.

Mean Motion in Stochastic Plasmas With a Space-Dependent Diffusion Coefficient

Radial transport in the tokamap, which has been proposed as a simple model for the motion in a stochastic plasma, is investigated. A theory for previous numerical findings is presented. The new results are stimulated by the fact that the radial diffusion coefficients is space-dependent. The space-dependence of the transport coefficient has several interesting effects which have not been elucidated so far. Among the new findings are the analytical predictions for the scaling of the mean radial displacement with time and the relation between the Fokker-Planck diffusion coefficient and the diffusion coefficient from the mean square displacement. The applicability to other systems is also discussed. (c) 2009 WILEY-VCH GmbH & Co. KGaA, Weinheim

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.

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.

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中的Kelvin-Helmholtz不稳定性

<正> Tokamak中的一个重要问题是加热。中性束注入加热是加热的一个有效手段,它使美国PLT上的离子温度达到7.1KeV.但PLT上的中性束注入的不对称性引起等离子体的快速环向旋转,转速可达1×10~7厘米/秒。1979年5月Suckewer等在PLT上测量了速度分布。 在具有速度剪切进行旋转的等离子体中,会不会形成新的磁流体力学不稳定性?1980年

Fusion energy-production from a deuterium-tritium plasma in the jet tokamak

Describes a series of experiments in the Joint European Torus (JET), culminating in the first tokamak discharges in deuterium-tritium fuelled mixture. The experiments were undertaken within limits imposed by restrictions on vessel activation and tritium usage. The objectives were: (i) to produce more than one megawatt of fusion power in a controlled way; (ii) to validate transport codes and provide a basis for accurately predicting the performance of deuterium-tritium plasmas from measurements made in deuterium plasmas; (iii) to determine tritium retention in the torus systems and to establish the effectiveness of discharge cleaning techniques for tritium removal; (iv) to demonstrate the technology related to tritium usage; and (v) to establish safe procedures for handling tritium in compliance with the regulatory requirements. A single-null X-point magnetic configuration, diverted onto the upper carbon target, with reversed toroidal magnetic field was chosen. Deuterium plasmas were heated by high power, long duration deuterium neutral beams from fourteen sources and fuelled also by up to two neutral beam sources injecting tritium. The results from three of these high performance hot ion H-mode discharges are described: a high performance pure deuterium discharge; a deuterium-tritium discharge with a 1% mixture of tritium fed to one neutral beam source; and a deuterium-tritium discharge with 100% tritium fed to two neutral beam sources. The TRANSP code was used to check the internal consistency of the measured data and to determine the origin of the measured neutron fluxes. In the best deuterium-tritium discharge, the tritium concentration was about 11% at the time of peak performance, when the total neutron emission rate was 6.0 × 10¹⁷ neutrons/s. The integrated total neutron yield over the high power phase, which lasted about 2 s, was 7.2 × 10¹⁷ neutrons, with an accuracy of ±7%. The actual fusion amplification factor, Q_DT was about 0.15

Jet:progress in performance and understanding

In 1990 JET operated with a number of technical improvements which led to advances in performance and permitted the carrying out of experiments specifically aimed at improving physics understanding of selected topics relevant to the "NEXT STEP". The new facilities include beryllium antenna screens, a prototype lower hybrid current drive system, and modification of the NI system to enable the injection of He-3 and He-4. Continued investigation of the hot-ion H-mode produced a value of n(D)(0)tau-E(T)(i)(0) = 9 x 10(20)m-3s keV, which is near conditions required for Q(DT) = 1, while a new peaked density profile H-mode was developed with only slightly lower performance. Progress towards steady state operation has been made by achieving ELMy H-modes under certain operating conditions, while maintaining good tau-E values. Experimental simulation of He ash transport indicates effective removal of alpha-particles from the plasma core for both L and H mode plasmas. Detailed analyses of particle and energy transport have helped establish a firmer link between particle and energy transport, and have suggested a connection between reduced energy transport and reversed shear. Numerical and analytic studies of divertor physics carried out for the pumped divertor phase of JET have helped clarify the key parameters governing impurity retention, and an intensive model validation effort has begun. Experimental simulation of alpha-particle effects with beta-fast up to 8% have shown that the slowing down processes are classical, and have given no evidence of deleterious collective effects.

Optimization of positions and currents of tokamak poloidal field coils using genetic algorithms

Plasma equilibrium geometry has a great influence on the confinement and magnetohydrodynamic stability in tokamaks. The poloidal field (PF) system of a tokamak should be optimized to support the prescribed plasma equilibrium geometry. In this paper, a genetic algorithm-based method is applied to solve the optimization of the positions and currents of tokamak PF coils. To achieve this goal, we first describe the free-boundary code EQT Based on the EQT code, a genetic algorithm-based method is introduced to the optimization. We apply this new method to the PF system design of the fusion-driven subcritical system and plasma equilibrium geometry optimization of the Experimental Advanced Superconducting Tokamak (EAST). The results indicate that the optimization of the plasma equilibrium geometry can be improved by using this method.

Propagation of the fast magnetosonic wave in a tokamak plasma

The propagation of the fast magnetosonic wave in a tokamak plasma has been investigated at low power, between 10 and 300 watts, as a prelude to future heating experiments.

The attention of the experiments has been focused on the understanding of the coupling between a loop antenna and a plasma-filled cavity. Special emphasis has been given to the measurement of the complex loading impedance of the plasma. The importance of this measurement is that once the complex loading impedance of the plasma is known, a matching network can be designed so that the r.f. generator impedance can be matched to one of the cavity modes, thus delivering maximum power to the plasma. For future heating experiments it will be essential to be able to match the generator impedance to a cavity mode in order to couple the r.f. energy efficiently to the plasma.

As a consequence of the complex impedance measurements, it was discovered that the designs of the transmitting antenna and the impedance matching network are both crucial. The losses in the antenna and the matching network must be kept below the plasma loading in order to be able to detect the complex plasma loading impedance. This is even more important in future heating experiments, because the fundamental basis for efficient heating before any other consideration is to deliver more energy into the plasma than is dissipated in the antenna system.

The characteristics of the magnetosonic cavity modes are confirmed by three different methods. First, the cavity modes are observed as voltage maxima at the output of a six-turn receiving probe. Second, they also appear as maxima in the input resistance of the transmitting antenna. Finally, when the real and imaginary parts of the measured complex input impedance of the antenna are plotted in the complex impedance plane, the resulting curves are approximately circles, indicating a resonance phenomenon.

The observed plasma loading resistances at the various cavity modes are as high as 3 to 4 times the basic antenna resistance (~ .4 Ω). The estimated cavity Q’s were between 400 and 700. This means that efficient energy coupling into the tokamak and low losses in the antenna system are possible.

Fluctuations in a tokamak plasma

An experimental investigation of low frequency floating potential fluctuations (f ≤ 200 kHz) in a research tokamak plasma using two spatially separated electrostatic probes has been performed. The spectra, correlation length, and the phase velocity of the fluctuations in both the radial and azimuthal direction have been determined. The propagation velocity in the toroidal direction was also measured and was found to be in the direction of electron current flow. The waves traveled azimuthally in the ion diamagnetic drift direction, even after the usual E x B rotation was taken into account. The electron density fluctuations associated with these oscillations were large, δn/n ≃ 0.35 - 0.50.

The spectra were found to have regularly spaced peaks which seemed to be related to specific azimuthal modes (m =1,2,3,...,etc. ) A parametric study was made to determine what effect plasma parameters had on these peaks. During periods of high electron density in the first 2 msec of the plasma lifetime, strong sawtooth type oscillations were observed. These oscillations typically had frequencies of approximately 10 kHz and were also present when large amounts of neutral gas were added during the discharge by a process called "gas puffing."

The results are compared with experimental observations made on other plasma devices with electric and magnetic probes and with microwave and CO₂ laser scattering techniques. (The scattering measurements are complimentary to the probe measurements since, in the former case, the wavelength is fixed by the scattering angle, but the oscillations could not be spatially localized.) The oscillations in the Caltech torus were probably related to a drift-tearing type instability which is thought to play a major role in the anomalous particle and energy flux observed in tokamaks. Comparisons are made between current theory and the experimental results. However, the theory for the observed oscillations is still in a rudimentary stage of development, and it is hoped that the present investigation will stimulate future analytical work.