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The behaviour of the scrape-off plasma of the projected Italian tokamak I-DTT is analysed by means of the two-dimensional edge code TECXY for either Li or Sn as the liquid target material. A scan in the outer midplane separatrix density range n e,omp ≈ (0.5-1.1) × 1020 m-3 is carried out for several power inputs into the scrape-off layer (SOL) up to a maximum of P SOL = 35 MW. The radiative capability and peak load mitigation are higher for Sn than for Li, but only slightly so at high density, so that the entire density range can be explored with liquid Sn targets, unlike Li whose employment at low density and high power is questionable. However, if a ceiling to the Sn concentration is set for high core fusion performance, Sn mitigation at low density may drop below Li. Nonetheless, concerns about the Li option derive from the very large material consumption and from the fact that the main mechanism of impurity release is sputtering, which is outside the full control of the operator. Considering specifically the three main scenarios-low, medium and high density-at the maximum auxiliary power of 45 MW, operations are successful with Sn, whereas Li requires an additional impurity able to radiate in the core, except at the highest density. The snowflake plus configuration, compared with the standard single null, as a sample for advanced divertors, appears more suited to exploit the advantages of the liquid metal option. Concerning additional impurities, only Ar and Ne are considered at present. They can strongly help in reaching and studying deep detached conditions even at the maximum power, but are not essential for operations. Wide flexibility in designing the target refrigeration system is required to fully exploit its great potential.

International audience

To understand the beneficial effects of impurity boron (B) on the plasma, the transport of B species in both atomic and ionic states in the edge plasma and the ensuing evolution of the background plasma are simulated during real-time B powder injection by using two-dimensional fluid code SOLPS-ITER. The initial background plasma profiles for the simulation are reconstructed based on an EAST upper-single null discharge (shot No. 70601). With the injection rate of B atoms, 4.0 × 1021 atoms/s, the simulation shows that effective charge number is less than 1.1 at the inner boundary R - Rsep = −5.2 cm (background plasma density, 5.1 × 1019 m−3) at the outer mid-plane indicating that the B impurity does not have much influence on the energy confinement. The present work finds that the impurity B exhibits very specific features in the edge plasma, mostly present in the narrow region just outside the separatrix. The particular distribution pattern can reduce the heat flux greatly, leading to the divertor detachment. The divertor detachment can be achieved easily under an injection rate of B powder in a reasonable range. This study suggests that the impurity B has a promising potential application in future tokamaks.

In this paper we present the comparison of simulations with the numerical codes COREDIV and SOLPS5.0 for JET L-mode discharges with ITER like wall (ILW). The simulations have been performed for L-mode shots with and without nitrogen seeding (#82291 - 9) which are characterised by relatively low auxiliary heating power (PNBI = 1.1 MW) and low electron density (ne = 2.35 × 1019 m–3). Comparisons are made to the experimental measurements (e.g. radiation levels, plasma profiles) and the differences between the results from the two codes (e.g. temperature and density profiles at the outer divertor plate) are shown and discussed.

ν eiωi]/c sωe), postulated to lead to shoulder growth through changes in SOL turbulence characteristics, correlates with increases in SOL shoulder amplitude, A s, but only under a subset of conditions (D2-fuelled L-mode density scans with outer strike point on the horizontal target). Λdiv fails to correlate with A s for cases of N2 seeding or during sweeping of the strike point across the horizontal target. The limited correlation of Λdiv and A s is also found for H-mode discharges. Thus, while it may be necessary for Λdiv to be above a threshold of ∼1 for shoulder formation and/or growth, another mechanism is required. More significantly, we find that in contrast to parallel resistivity, outer divertor recycling, as quantified by the total outer divertor Balmer D α emission, I-D α, does scale with A s where Λdiv does and even where Λdiv does not. Divertor recycling could lead to SOL density shoulder formation through: (a) reducing the parallel to the field flow (loss) of ions out of the SOL to the divertor; and (b) changes in radial electric fields which lead to E × B poloidal flows as well as potentially affecting SOL turbulence birth characteristics. Thus, changes in divertor recycling may be the sole process involved in bringing about SOL density shoulders or it may be that it acts in tandem with parallel resistivity. © EURATOM 2018., The low temperature boundary layer plasma (scrape-off layer or SOL) between the hot core and the surrounding vessel determines the level of power loading, erosion and implantation of material surfaces, and thus the viability of tokamak-based fusion as an energy source. This study explores mechanisms affecting the formation of flattened density profiles, so-called 'density shoulders', in the low-field side (LFS) SOL, which modify ion and neutral fluxes to surfaces - and subsequent erosion. There is evidence against local enhancement of ionization inducing shoulder formation. We find that increases in SOL parallel resistivity, Λdiv (=[L

The impact of the finite grid size in SOLPS-ITER edge plasma simulations is assessed for JET H-mode discharges with a metal wall. For a semi-horizontal divertor configuration it is shown that the separatrix density is at least 30% higher when a narrow scrape-off layer (SOL) grid width is chosen in SOLPS-ITER compared to the case for which the SOL grid width is maximised. The density increase is caused by kinetic neutrals being not confined inside the divertor region because of the reduced extent of the plasma grid. In this case, an enhanced level of reflections of energetic neutrals at the low-field side (LFS) metal divertor wall is observed. This leads to a shift of the ionisation source further upstream which must be accounted for as a numerical artefact. An overestimate in the cooling at the divertor entrance is observed in this case, identified by a reduced heat flux decay parameters lambda(div)(q). Otherwise and further upstream the mid-plane heat decay length lambda(q) parameter is not affected by any change in divertor dissipation. This confirms the assumptions made for the ITER divertor design studies, i.e. that lambda(q) upstream is essentially set by the assumptions for the ratio radial to parallel heat conductivity. It is also shown that even for attached conditions the decay length relations lambda(ne)>lambda(Te)>lambda(q) hold in the near-SOL upstream. Thus for interpretative edge plasma simulations one must take the (experimental) value of lambda(ne) into account, rather than lambda(q), as the former actually defines the required minimum upstream SOL grid extent. For complete list of authors see http://dx.doi.org/10.1016/j.nme.2018.10.013

The design of a thermonuclear fusion reactor, including ITER, requires a detailed description of the confined plasma properties (confinement time, radiation, stability, etc.) as well as its interaction with the reactor wall (mainly the divertor). Indeed, the operational space of a fusion reactor is strongly limited by the amount of power per surface unit that the wall material can extract locally. With respect to the symmetries of the system, the heat load can be parameterized by two quantities: a peak value and a spatial width. Currently, there is no physical model able to predict the peak value and the width of this heat load. Empirical scaling laws seem to indicate certain trends, but they do not allow estimating precisely the impact on the operational domain of ITER. This uncertainty is the result of several aspects of this physics: (i) the physics of heat and particle transport in the region at the interface between the divertor and the confined plasma (the scrape-off layer) is still not well understood, (ii) measurements are complex and their interpretation is strongly linked to the nature of transport, (ii) very few models bring a physical understanding to experiment trends. The PhD project aims at re-establishing the analyses without separating physical models from measure interpretation and experimental signatures. Therefore this approach is based on both the consistency of global understanding and physical meaning of results. (a) Physics of the measure and uncertainties. Accounting for errors and uncertainties in the interpretation of experimental measurements is quite delicate, due to self-organization properties and the strong inhomogeneity of this plasma region. The analysis of simple experimental conditions should allow estimating uncertainties and confidence intervals in the trends. (b) Analysis of data bases and scaling laws. An analysis should be developed on the basis of constrains defined by the physics, by using dimensionless control parameters. A coherent approach of the interpretations across diverse experimental conditions will be developed from the analyses of data issued by the integrated modelling code SOLEDGE2D. (c) Theoretical analysis on the nature of transport in this plasma region. Theoretical models have to guide the experimental analysis, which in turns validates or invalidates the tested hypothesis. It seems particularly important to consider the role of very strong fluctuations (plasma turbulence) and the link with transport in the confined region, for instance with the large scale flows., Le dimensionnement d'un réacteur à fusion thermonucléaire, incluant ITER, nécessite une description détaillée des propriétés du plasma confiné (temps de confinement, rayonnement, stabilité, etc.) ainsi que de son interaction avec les parois du réacteur. En effet, l'espace opérationnel d'un réacteur à fusion est fortement limité par la densité surfacique de puissance que l'on peut extraire localement. Compte tenu des symétries, le dépôt d'énergie peut être caractérisé par deux grandeurs: la valeur pique et une extension spatiale. A l'heure actuelle, il n'existe pas de modèles physiques susceptibles de prédire la largeur et la valeur pique de ce dépôt. Les lois d'échelle empiriques semblent indiquer certaines tendances mais ne permettent pas d'estimer avec précision l'impact sur le domaine opérationnel d'ITER. Cette marge d'incertitude est reliée à plusieurs points délicats de cette physique (i) la physique du transport de la chaleur et des particules dans la région à l'interface entre divertor et plasma confiné (la SOL) est encore mal comprise, (ii) les mesures sont complexes et leur interprétation fortement liée à la physique du transport, (iii) très peu de modèles parviennent à donner un sens physique aux lois d'échelle. Le projet de thèse vise à refonder l'analyse sans séparer les modèles physique de l'interprétation des mesures et des signatures expérimentales. Cette approche est donc basée sur la cohérence de la compréhension et le sens physique des résultats. Physique de la mesure et incertitudes. La prise en compte des erreurs et des incertitudes dans l'interprétation des mesures expérimentales est assez délicate en raison des propriétés d'auto-organisation et de forte inhomogénéité de cette région du plasma. L'analyse en partant de situations expérimentales simples permettra d'estimer l'incertitude et la marge de confiance dans les tendances. Analyse des bases de données et lois d'échelle. Il conviendra de développer une analyse sur la base des contraintes sur la physique en utilisant les paramètres de contrôle sans dimension. En utilisant le code de modélisation intégrée SOLEDGE2D, il conviendra de resituer les mesures dans la diversité des situations expérimentales en s'assurant de la cohérence avec l'ensemble des mesures existantes. Analyse théorique sur la nature du transport sur cette région. Les modèles théoriques doivent servir de guide à l'analyse expérimentale qui doit valider ou invalider les hypothèses qui sont faites. Il semble particulièrement important de dégager le rôle des très fortes fluctuations (turbulence plasma) et les liens avec le transport dans la région confinée, tels que les écoulements à grande échelle.

The design of a thermonuclear fusion reactor, including ITER, requires a detailed description of the confined plasma properties (confinement time, radiation, stability, etc.) as well as its interaction with the reactor wall (mainly the divertor). Indeed, the operational space of a fusion reactor is strongly limited by the amount of power per surface unit that the wall material can extract locally. With respect to the symmetries of the system, the heat load can be parameterized by two quantities: a peak value and a spatial width. Currently, there is no physical model able to predict the peak value and the width of this heat load. Empirical scaling laws seem to indicate certain trends, but they do not allow estimating precisely the impact on the operational domain of ITER. This uncertainty is the result of several aspects of this physics: (i) the physics of heat and particle transport in the region at the interface between the divertor and the confined plasma (the scrape-off layer) is still not well understood, (ii) measurements are complex and their interpretation is strongly linked to the nature of transport, (ii) very few models bring a physical understanding to experiment trends. The PhD project aims at re-establishing the analyses without separating physical models from measure interpretation and experimental signatures. Therefore this approach is based on both the consistency of global understanding and physical meaning of results. (a) Physics of the measure and uncertainties. Accounting for errors and uncertainties in the interpretation of experimental measurements is quite delicate, due to self-organization properties and the strong inhomogeneity of this plasma region. The analysis of simple experimental conditions should allow estimating uncertainties and confidence intervals in the trends. (b) Analysis of data bases and scaling laws. An analysis should be developed on the basis of constrains defined by the physics, by using dimensionless control parameters. A coherent approach of the interpretations across diverse experimental conditions will be developed from the analyses of data issued by the integrated modelling code SOLEDGE2D. (c) Theoretical analysis on the nature of transport in this plasma region. Theoretical models have to guide the experimental analysis, which in turns validates or invalidates the tested hypothesis. It seems particularly important to consider the role of very strong fluctuations (plasma turbulence) and the link with transport in the confined region, for instance with the large scale flows., Le dimensionnement d'un réacteur à fusion thermonucléaire, incluant ITER, nécessite une description détaillée des propriétés du plasma confiné (temps de confinement, rayonnement, stabilité, etc.) ainsi que de son interaction avec les parois du réacteur. En effet, l'espace opérationnel d'un réacteur à fusion est fortement limité par la densité surfacique de puissance que l'on peut extraire localement. Compte tenu des symétries, le dépôt d'énergie peut être caractérisé par deux grandeurs: la valeur pique et une extension spatiale. A l'heure actuelle, il n'existe pas de modèles physiques susceptibles de prédire la largeur et la valeur pique de ce dépôt. Les lois d'échelle empiriques semblent indiquer certaines tendances mais ne permettent pas d'estimer avec précision l'impact sur le domaine opérationnel d'ITER. Cette marge d'incertitude est reliée à plusieurs points délicats de cette physique (i) la physique du transport de la chaleur et des particules dans la région à l'interface entre divertor et plasma confiné (la SOL) est encore mal comprise, (ii) les mesures sont complexes et leur interprétation fortement liée à la physique du transport, (iii) très peu de modèles parviennent à donner un sens physique aux lois d'échelle. Le projet de thèse vise à refonder l'analyse sans séparer les modèles physique de l'interprétation des mesures et des signatures expérimentales. Cette approche est donc basée sur la cohérence de la compréhension et le sens physique des résultats. Physique de la mesure et incertitudes. La prise en compte des erreurs et des incertitudes dans l'interprétation des mesures expérimentales est assez délicate en raison des propriétés d'auto-organisation et de forte inhomogénéité de cette région du plasma. L'analyse en partant de situations expérimentales simples permettra d'estimer l'incertitude et la marge de confiance dans les tendances. Analyse des bases de données et lois d'échelle. Il conviendra de développer une analyse sur la base des contraintes sur la physique en utilisant les paramètres de contrôle sans dimension. En utilisant le code de modélisation intégrée SOLEDGE2D, il conviendra de resituer les mesures dans la diversité des situations expérimentales en s'assurant de la cohérence avec l'ensemble des mesures existantes. Analyse théorique sur la nature du transport sur cette région. Les modèles théoriques doivent servir de guide à l'analyse expérimentale qui doit valider ou invalider les hypothèses qui sont faites. Il semble particulièrement important de dégager le rôle des très fortes fluctuations (turbulence plasma) et les liens avec le transport dans la région confinée, tels que les écoulements à grande échelle.

V simulacijah plasti postrgane plazme (SOL) z B2-EIRENE oziroma SOLPS (Scrape-Off Layer Plasma Simulations) se pojavlja veliko število vhodnih parametrov, možnih kombinacij ionskih delcev, atomskih procesov in interakcij med plazmo in steno. Simulacije SOLPS se izvajajo na superračunalnikih v uporabniških imenikih in se lahko shranjujejo tudi v podatkovne baze MDSPLUS ali pretvorijo za zapis v fuzijske podatkovne baze s predpisanimi strukturami za integrirano modeliranje (EUROfusion CPO ali ITER IDS). To magistrsko delo raziskuje podatkovno strukturo EUROfusion CPO in ITER IDS ter pridobljene informacije uporabi za pripravo primernih orodij za transformacijo podatkov iz CPO v IDS ter izdelavo vtičnika in nadgradenj splošno-namenskega vizualizacijskega orodja ParaView, ki bo omogočal primerjalne analize rezultatov simulacij SOLPS. Nenazadnje je bila nadgrajena koda SOLPS-ITER, ki omogoča zapis edge_profiles IDS-jev v podatkovno bazo IMAS in tako nudi nabor orodij za okolje integriranega modeliranja. In Scrape-Off Layer Plasma Simulations (SOLPS) with B2-EIRENE there are many input parameters, possible combinations of ion particles, atomic processes, and interactions between the edge plasma and the core boundary and plasma-facing components. The SOLPS are carried out on a cluster in a user directory and can be stored in the MDSPLUS database or can be converted to a "record" in fusion databases with prescribed structures for integrated modeling (EUROfusion CPO or ITER IDS). This thesis investigates the data structure of EUROfusion CPO and ITER IDS and uses gathered information to create appropriate tools for data transformation from CPO to IDS together with plugins and upgrades for general purpose visualization tool ParaView which allows comparative analysis of SOLPS output results. Finally, the SOLPS-ITER code is upgraded to write edge_profiles IDS into IMAS database providing a required set of tools for in integrated modeling framework.

The term “shoulder formation” refers to an increase of the density decay length in the scrape-off layer (SOL) observed in many tokamaks during L-mode operation when a density threshold is reached. Recent experiments in ASDEX Upgrade (AUG) and JET have shown that the shoulder forms when the divertor collisionality in the divertor electrically disconnects filaments from the wall. This leads to a transition from the sheath limited to the inertial regime and to an enhancement of radial particle transport, in good agreement with analytical models. In the present work, the validity of such a mechanism is investigated in the more reactor-relevant H-mode regime. For this, a cold divertor H-mode scenario is developed in AUG using different levels of D puffing and N seeding, in which inter-ELM filaments and SOL density profiles are measured. The basic relation between filament size and divertor collisionality is still valid in H-mode plasmas, albeit an additional condition related to the gas fueling rate has been found for the formation of the shoulder. PSI-20 keywords: ASDEX-Upgrade, Detachment, Edge plasma, Intermittent transport, H-mode

A series of neon seeded JET ELMy H-mode pulses is considered from the modeling as well as from the experimental point of view. For two different Ne seeding rates and two different D puffing gas levels the heating power, P-heat, is in the range 22-29.5 MW. The main focus is on the numerical reconstruction of the total radiated power (which mostly depends on the W concentration) and its distribution between core and divertor and of Z(eff)(which mostly depends on the Ne concentration). To model self-consistently the core and the SOL two input parameters had to be adjusted case by case: the SOL diffusivity, D SOL, and the core impurity inward pinch, v(pinch). D-SOL had to be increased with increasing Gamma(Ne) and the level of v(pinch) had to be changed, for any given Gamma(Ne), according to the level of P-heat : it decreases with increasing P-heat. Since the ELM frequency, f(ELM), is experimentally correlated with P-heat, (it increases with P-heat) the impurity inward pinch can be seen as to depend on f(ELM). Therefore, to maintain a low v(pinch) level (i.e. high f(ELM)) Gamma(Ne)/P-heat should not exceed a certain threshold, which slightly increases with the Gamma(D) puffing rate. This might lead to a limitation in the viability of reducing the target heat load by Ne seeding at moderate Gamma(D), while keeping Z(eff) at acceptably low level. (C) 2016 Published by Elsevier Ltd. For complete list of authors see http://dx.doi.org/10.1016/j.nme.2016.10.012

International audience; In order to improve cross-field transport description, drifts and currents have been implemented in SOLEDGE2D-EIRENE. The derivation of an equation for the electric potential is recalled. The resolution of current equation is tested in a simple slab case. WEST divertor simulations in forward-B and reverse-B fields are also discussed. A significant increase of ExB shear is observed in the forward-B configuration that could explain a favorable L-H transition in this case.

Recent JET (ITER-Like Wall) experiments have shown that the fueling gas puffed from different locations of the vessel can result in different scrape-off layer (SOL) density profiles and therefore different radio frequency (RF) coupling. To reproduce the experimental observations, to understand the associated physics and to optimize the gas puff methods, we have carried out three-dimensional (3D) simulations with the EMC3-EIRENE code in JET-ILW including a realistic description of the vessel geometry and the gas injection modules (GIMs) configuration. Various gas puffing methods have been investigated, in which the location of gas fueling is the only variable parameter. The simulation results are in quantitative agreement with the experimental measurements. They confirm that compared to divertor gas fueling, mid-plane gas puffing increases the SOL density most significantly but locally, while top gas puffing increases it uniformly in toroidal direction but to a lower degree. Moreover, the present analysis corroborates the experimental findings that combined gas puff scenarios—based on distributed main chamber gas puffing—can be effective in increasing the RF coupling for multiple antennas simultaneously. The results indicate that the spreading of the gas, the local ionization and the transport of the ionized gas along the magnetic field lines connecting the local gas cloud in front of the GIMs to the antennas are responsible for the enhanced SOL density and thus the larger RF coupling. EURATOM 633053

In this paper we present the comparison of simulations with the numerical codes COREDIV and SOLPS5.0 for JET L-mode discharges with ITER like wall (ILW). The simulations have been performed for L-mode shots with and without nitrogen seeding (#82291 - 9) which are characterised by relatively low auxiliary heating power (PNBI = 1.1 MW) and low electron density (ne = 2.35 × 1019 m–3). Comparisons are made to the experimental measurements (e.g. radiation levels, plasma profiles) and the differences between the results from the two codes (e.g. temperature and density profiles at the outer divertor plate) are shown and discussed.

Zuk��nftige Kernfusionsanlagen wie ITER und DEMO werden aller Wahrscheinlichkeit nach in der H-mode (high confinement mode) betrieben. Durch das Auftreten einer Randtransportbarriere (ETB) weist dieser Betriebsmodus hohe Druck- und Stromgadienten am Plasmarand auf, welche freie Energie f��r magnetohydrodynamische Instabilit��ten liefern. Eine Eigenschaft der H-mode ist das quasi periodische Auftreten von sogenannten ELMs (Edge Localized Modes) auf, die zur Zerst��rung der ETB f��hren. Der Mechanismus, der die kritischen Gro��en die zur Instabili��t eines ELMs f��hren momentan am besten beschreibt ist das Peeling-Balloning Modell, in dem Plasmastrom getriebene Peeling Moden mit Druck getriebenen Ballooning Moden zu instabilen Peeling-Ballooning Moden koppeln. Diese Masterarbeit besch��ftigt sich mit der Untersuchung nichtlinearer Kopplung magnetischer Moden in ASDEX Upgrade Plasmaentladungen. In einem theoretischen Modell wird erst gezeigt, wie es zu Kopplungen von Moden in Wellen und demanach auch in Plasmafluktuationen kommen kann. In der Bikoher��nzanalyse wird eine Methode diese Kopplungen zu diagnostizieren gezeigt. Ein Bikoh��renz Algorithmus wurde entwickelt um beliebige zeitlich aufgel��ste Signale zu analysieren. Mit der Anwendung von Fensterfunktionen und Dithering wurden zwei Methoden gefunden eine Verbesserung der Sensitivit��t bei Hintergrundrauschen und der Stabilit��t bei numerischen Fehlern zu erreichen. Um die Funktion des Algorithmus bei der Analyse experimenteller Daten zu zeigen, wurden Signale von magnetischen Pick-up Spulen aus verschiedensten AUG Entladungen analysiert und dargestellt. Um nun die Kopplung verschiedener Moden vor, w��hrend und nach den ELMs zu untersuchen wurden die magnetischen Signale an die ELM Anf��nge synchronisiert. Mit dieser Methode bleiben die ausschlaggebenden Eigenschaften der ELMs erhalten w��hrend sich unwichtige Dinge wegmitteln. Die somit erhaltenen ELM-synchronisierten Daten wurden dann in eindeutige Phasen unterteilt und getrennt mit der Bikoh��renzanalyse untersucht, wobei keine offensichtlichen Modenkopplungen in den einzelnen ELM Phasen gefunden wurden., The forseen operation scenario for future fusion devices like ITER and DEMO is the high confinement mode (H-mode), which is characterized by high pressure and current density gradients in the plasma edge due to an edge transport barrier (ETB). These steep gradients provide a source of free energy for magnetohydr�� dynamic instabilities. One property of the H-mode are the quasi periodically occuring edge localized modes (ELMs), which lead to the degradation of the ETB. The best model so far, which describes the critical quantities driving the ELM unstable, is the peeling-ballooning model, in which peeling modes, driven by the current density and its gradients, as well as ballooning modes, driven by the pressure gradients, combine to form peeling-ballooning modes. This Master Thesis deals with non-linear coupling of magnetic modes in AS- DEX Upgrade plasma discharges. A theoretical model first shows how coupling of modes in waves, and therefore also in plasma fluctuations, develops. Bicoherence analysis delivers a method to diagnose and dispaly these couplings. Over the course of this thesis a bicoherence algorithm was developed and thoroughly tested. Windowing and dithering turned out to be useful tools to decrease the noise sensitivity and increase the stability of the algorithm. To test the capabilities of the algorithm in analyzing experimental data, signals of magnetic pick-up coils from different AUG dischages were anayzed and displayed. The magnetic signals were then synchronized to the ELM onsets to check for coupling before, during and after the ELMs. This method preserves the important properties of the ELMs while removing unimportant ones. Consequently these ELM synchronized signals were divided into distinct phases and separatly analyzed with the bicoherence algorithm. No apparent mode couplings were found in any of the ELM phases.

Zuk��nfitge Kernfusionsanlagen wie ITER und DEMO sollen in der H-mode (high confinement mode) betrieben werden. Dieser Betriebsmodus zeichnet sich durch hohe Druck-, Temperatur- und Dichtegadienten am Plasmarand aufgrund des Auftretens der Randtransportbarriere (ETB) aus. Dies f��hrt wiederum zu erh��htem Druck, Temperatur und Dichte im Inneren des Plasmas. Jedoch treten in diesem Betriebsmodus sogenannte ELMs (edge localized modes) auf, die zur Zerst��rung der ETB und dadurch zu flachen Gradienten am Plasmarand f��hren. Dies hat zur Folge, dass Teilchen aus dem Plasmarand auf die Gef����wand und die Divertoren fliegen. Das Verst��ndnis der darauf folgenden Entwicklung bis zur Wiederherstellung der ETB ist von gro��er Bedeutung, da diese Einsicht in die dominanten Transportprozesse in der ETB gibt. Im Rahmen dieser Arbeit wurde die ETB und das zeitaufgel��ste Verhalten des inneren und ��u��eren Divertors bei unterschiedlichen Deuterium Neutralgasraten untersucht. Dabei konnte gezeigt werden, dass die Entwicklung der Elektronendichte und -temperatur an der Mittelebene nach dem ELM mit der Dichte- und Temperaturentwicklung am ��u��eren Divertor und mit Fluktuationen an der Mittelebene zusammenh��ngt. Eine deutliche Korrelation zwischen der Dichte- und Temper�� turentwicklung des ��u��eren und inneren Divertors ist nicht zu erkennen. Die simultane, ELM synchronisierte Betrachtung der Mittelebe, des ��u��eren Divertors und der Fluktuationen f��hren zu einem Modell, bei dem die Rate des Dichteanstiegs an der Mittelebene als Balance des Zwischenspiels von Teilchenaussto�� durch magnetische Turbulenzen und einer lokalen Ionisationsquelle innerhalb der Separatrix gesehen werden kann. Dabei kann der zweite Elektronendichteanstieg am ��u��eren Divertor durch das Auftreten von Fluktuationen an der Mittelebene in dieser Arbeit erkl��rt werden. Die Ionisationsquelle, die von Teilchen herr��hrt, die aufgrund von Recycling an der Gef����wand auftreten, steigt nach dem ELM an und kann mit dem KN1D Code modelliert werden. Abschlie��end f��hrt diese Modellierung zusammen mit der Kontinuit��tsgleichung, welche auf dem Drift-Diffusions-Modell basiert, zu einer Absch��tzung des von Fluktuationen stammenden, zus��tzlichen Teilchentransports in der Gr����enordnung von 1 m2/s, Future fusion devices like ITER and DEMO will be operated in the high confinement mode (H-mode). The H-mode is characterized by high pressure, temperature and density gradients in the plasma edge due to an edge transport barrier, and therefore high pressure, temperature and density in the plasma core. However, so called edge localized modes (ELMs) lead to the destruction of the edge transport barrier and particles and heat are expelled onto the plasma facing components, e.g. the vessel wall and the divertor. The post-ELM edge plasma density and temperature evol�� tions are of major interest as they shed light on the dominant transport processes in the edge transport barrier. In this thesis the inter-ELM evolution of the edge transport barrier is investigated together with the time-resolved behaviour of outer and inner divertor for different gas puffing rates. It is shown that the low field side inter-ELM evolution of the midplane density and temperature is linked to the outer divertor density and temperature and to midplane fluctuations at the plasma edge. A clear link between the outer and inner divertor inter-ELM evolution cannot be identified. The simultaneous examination of the inter-ELM evolution of the midplane edge profiles, outer divertor profiles and midplane fluctuations leads to a model wherein the density recovery rate can be seen as a result of an interplay between a local ionization source, due to incoming neutrals originating from the chamber wall, and the outward particle flux, due to particle transport caused by fluctuations seen as magnetic oscillations. The transient density increase at the outer divertor due to an ELM can be followed by a second density rise. The mechanism of this phenomenon can be explained through the occurrence of midplane fluctuations. The ionization source, due to incoming neutrals, which are recycled from plasma facing components, increases after an ELM event and is modelled using a one dimensional transport algorithm KN1D. This transport algorithm and the continuity equation, assuming a drift-diffusion model, lead to the estimation of additional particle transport, that originates from midplane fluctuations, of the order of 1 m2/s

The transport of heat and particles in the edge of tokamaks plays a key role in both the performance of the confined plasma and the extraction of power and thus the lifetime of the plasma facing components. It’s in this context that this thesis is inscribed, which focuses on the role played by the transverse magnetic field flows in the balance between parallel and perpendicular dynamic that governs the edge region of a tokamak. These flows can produce poloidal asymmetries of heat and particles deposit on plasma facing components and generally asymmetries of various amounts in plasma. The radial drift velocities are due to the presence of a radial electric field resulting from charge balance (electric drift velocity) or related to effects of the toroidal geometry inducing a magnetic field inhomogeneity (curvature drift velocity). To advance the understanding of these phenomena, numerical modeling of transport and turbulence in complex geometries is essential. In addition, synthetic diagnostic tools for modeling the measurement process in numerical plasmas are developed to enable a realistic comparison between models and experiments. Modeling of perpendicular drift velocities was introduced into the SOLEDGE2D code describing the transport of the density, momentum and energy of a tokamak plasma. We first studied the impact of a prescribed electric field on plasma equilibrium to understand the mechanisms behind plasma asymmetries and study the establishment of parallel flows and asymmetry of the heat flux on plasma facing components. Then we implemented a self-consistent model solving the electric potential in SOLEDGE2D fluid equations to understand the equilibrium of the electric field and to study the effect of the magnetic configuration of the tokamak and the curvature drift velocity on it. In the second part of this thesis, a synthetic diagnosis modeling the experimental measurements of Doppler backscattering was developed and tested in order to be applied to simulations of 3D turbulent fluid code TOKAM3X. This diagnosis measures the perpendicular velocity of the plasma from the movement of the density fluctuations. It was used to compare the perpendicular velocity asymmetries observed experimentally to asymmetries measured in numericalsimulations., Le transport des particules et de la chaleur dans la zone de bord des tokamaks joue un rôle déterminant à la fois sur les performances du plasma confiné et sur l’extraction de la puissance et ainsi la durée de vie des composants face au plasma. C’est dans ce contexte que s’inscrit ce travail de thèse, qui porte sur le rôle joué par les écoulements transverses au champ magnétique dans l’équilibre entre dynamique parallèle et dynamique perpendiculaire qui régit la région périphérique d’un tokamak. Ces écoulements peuvent produire des asymétries poloïdales du dépôt de chaleur et de particules sur les composants face au plasma, et plus généralement des asymétries des diverses quantités dans le plasma. Les vitesses de dérive radiale sont d’origine électrique (liées à la présence d’un champ électrique radial résultant de l’équilibre des charges) ou liées aux effets de la géométrie toroïdale induisant une inhomogénéité du champ magnétique (vitesse de gradient-courbure). Pour progresser dans la compréhension de ces phénomènes, la modélisation numérique du transport et de la turbulence en géométrie complexe est indispensable. En complément, des outils de diagnostic synthétique permettant de modéliser les processus de mesure dans les plasmas numériques sont développés pour permettre une comparaison réaliste entre modèles et expériences. La modélisation des vitesses de dérive perpendiculaire a été introduite dans le code SOLEDGE2D décrivant le transport de la densité, quantité de mouvement et énergie d’un plasma de tokamak. Nous avons d’abord étudié l’impact d’un champ électrique prescrit sur les équilibres plasma, pour comprendre les mécanismes à l’origine des asymétries du plasma et étudier l’établissement d’écoulement parallèle et d’asymétrie du dépôt de chaleur sur les composants face au plasma. Nous avons ensuite implémenté un modèle auto-consistant de résolution du potentiel électrique dans les équations fluides de SOLEDGE2D afin de comprendre l’équilibre du champ électrique et d’étudier l’effet de la configuration magnétique du tokamak et de la vitesse de gradient-courbure sur ce dernier. Dans la deuxième partie de cette thèse, un diagnostic synthétique permettant de modéliser les mesures expérimentales de rétro-diffusion Doppler a été développé et testé en vue d’être appliqué aux simulations du code fluide 3D turbulent, TOKAM3X. Ce diagnostic permet de mesurer la vitesse perpendiculaire du plasma à partir du mouvement des fluctuations de densité. Il a été utilisé ici pour comparer les asymétries de vitesse observées expérimentalement aux asymétries mesurées dans les simulations numériques.

It is important to obtain spatial structure of plasma and neutral particles near plasma facing material in order to understand plasma dynamic and recycling phenomena in divertor region of magnetically plasma confinement device. In the compact PWI simulator APSEDAS, the ratio of helium visible lines at 667.8 nm and at 728.1 nm is measured to estimate electron density near tungsten target. H_α intensity is also obtained from measurement using CCD camera. It is found that both the electron density and the H_α intensity show similar distribution in which both parameters decrease toward the tungsten target. Based on the Monte Carlo simulation, the H_α intensity due to the neutral particle from filling gas source sufficiently shows similar tendency with that captured by CCD camera and the electron density.

H_α emission images near metal surface were observed based on a CCD camera measurement in the Advanced PWI Simulation Experimental Device and Analysis System (APSEDAS). In the axial direction of cylindrical plasma, it was found that the H_α emission gradually decrease toward the metal surface and drastically drop in the vicinity of the metal surface. It suggests that the axial profile of the H_α emission mainly reflect on that of electron density. The H_α emission also showed different dependence between at inductively-coupled plasma arid at helicon wave plasma in the experiments of hydrogen filling pressure scan. It was clarified that the H_α emission in the case of inductively-coupled plasma represented a more significant decrease toward the metal surface with the increase of the filling pressure.

IGNITOR is the only experiment today designed with the ambitious goal of achieving ignition. During the discharge, the First-Wall Limiter (FWL) will receive about 20 MW [B. Coppi, A. Airoldi, F. Bombarda, et al., Nucl. Fusion 41 (200 1) 1253-1257] power, split between conduction/convection and radiation channels. Previous estimates suggested a peak heat flux on the FWL close to 1 MW/m(2). This value should be re-evaluated, accounting for the latest developments in the design of the IGNITOR first wall and changes in the operational scenarios, in order to assess the risk of damages to the FWL deriving from the combination of thermal and mechanical stresses. For this purpose, the development of the ASPOEL code, implementing a 2D model of the edge plasma and including a detailed representation of the FWL, was started at Politecnico di Torino. Here we present the main features of the new code, and illustrate its potential via a preliminary application to IGNITOR. (c) 2007 Elsevier B.V. All rights reserved.

A novel probe and approach to the direct measurements of the plasma potential in a strong magnetic field is suggested. The principle of this method is to reduce the electron saturation current to the same magnitude as that of the ion saturation current. In this case, the floating potential of the probe becomes indentical to the plasma potential. This goal is attained by a shield, which screens off an adjustable part of the electron current from the probe collector due to the much smaller gyro-radius of the electrons. First systematic measurements have been perfomred in the CASTOR tokamak.

The response of the edge plasma to a magnetic perturbation (MP) is studied in the RFX-mod device, in both reversed field pinch (RFP) and tokamak discharges. The use of spatially distributed diagnostics, in particular along the poloidal direction, allows a direct 3D characterization of the plasma. It is shown that there is a difference between the spectrum of the MP and that of different properties of the edge plasma, such as the floating potential, the electron pressure and the flow, for both RFP and tokamak plasmas. In particular, for RFP cases, even if the magnetic perturbation is mainly m = 1, with the amplitude of the m = 0 and m = 2 sidebands negligible, the connection length to the wall of the magnetic field L cw has a more complex structure. The ergodic regions, with L cw larger than the Kolmogorov length, are due to the interaction of both m = 1 and m = 0 modes, showing a complex and non-monochromatic behaviour along the poloidal angle. This structure of L cw is responsible for the impure m = 1 behaviour of the plasma wall interaction, floating potential and electron pressure. Thus the helicity of the edge plasma is quite different from that of the dominant MP.

A reciprocating probe head with six pins was used for localized measurements of electric fields and densities in the scrape-off layer (SOL) of ASDEX Upgrade (AUG) up to the edge shear layer (SL) near the Last Closed Flux Surface (LCFS). The edge SL is characterized by a strong sudden change in the poloidal velocity v close to the separatrix. The probes were used to determine this velocity by different methods which are critically compared to each other concerning their reliability. By the first method the poloidal velocity was deduced from the radial electric field E-r measured by two radially staggered probe pins, with v being due to the E-r x B-phi-drift (B-phi is the toroidal field). The two other methods utilized the cross correlation of two poloidally staggered ion-biased probes and two poloidally staggered floating probes, respectively. In this case the time lags with maximum cross correlation were used to determine the poloidal velocity and its jump, yielding comparable results to the first method. Also the method of conditional averaging was applied to the latter signals.