Pour réaliser les conditions des réactions de fusion thermonucléaire dans le tokamak ITER, des moyens additionnels de chauffage sont requis. L'une des principales méthodes pour chauffer les ions du plasma de coeur sera l'injection de neutres D0 énergétiques. Le neutraliseur est l'étape de l'injecteur de neutres d'ITER où le faisceau de deutérium prend ses propriétés en termes de taux de neutres D0 et de direction de propagation. L'interaction entre le faisceau à 1MeV et le gaz D2 neutralisant (~0.1Pa) crée un plasma secondaire. Les phénomènes physiques en jeu sont présentés à travers l'analyse des résultats du code OBI-2. OBI-2 est un code PIC-MCC (Particle In Cell Monte Carlo Collision) en géométrie cylindrique (2D3V) développé au LPGP qui permet de suivre la propagation du faisceau et les particules du plasma le long du neutraliseur.L'injection de lithium comme cible neutralisante a été étudiée et comparée au deutérium. Une étude paramétrique sur le neutraliseur basé sur le lithium a été réalisée dans la mesure où la longueur et/ou la densité de Li injectée peuvent être modifiées. Le profil de densité de Li a été estimé par le code Monte-Carlo 3D MC-OLIJET développé au LPGP. Le profil résultatnt a été implémenté en entrée du code PIC-MCC. Les résultats montrent la faisabilité du neutraliseur basé sur le lithium, gardant la convergence correcte du faisceau et avec de meilleures performances en termes de durée de vie des cryompompes avant régénération, de neutralisation du faisceau, d'effet de rétrodiffusion des ions positifs.

The neutron spectrometer TOFOR was installed at JET in 2005 for high-rate observation of neutrons from reactions between two deuterium (D) ions. Neutron spectrometry as a fusion plasma diagnostic technique is invoked to obtain information about the velocity states of fusion fuel ions. Based on neutron spectrometry data, conclusions can be drawn on the efficiency of plasma heating schemes as well as optimization of fuel ion confinement. The quality of TOFOR analysis is found to depend on how well the instrument response function is known; discriminator threshold levels, detector time alignment and electronics broadening are identified as crucial issues. About 19 percent of the neutrons observed with TOFOR have scattered off the JET vessel wall or other structures in the line-of-sight before reaching the instrument, as established through simulations and measurements. A method has been developed to take these neutrons into account in the analysis. TOFOR measurements of fast deuterium distributions are seen to agree with distributions deduced from NPA data, obtained based on an entirely different principle. This serves as validation of the modeling and analysis. Extraordinary statistics in the TOFOR measurements from JET pulses heated with 3rd harmonic RF heating on D beams allow for study of instabilities using neutron emission spectrometry. At ITER, similar studies should be possible on a more regular basis due to higher neutron rates. Observations of neutrons from Be+3He reactions in the TOFOR spectrum from D plasmas heated with fundamental RF tuned to minority 3He raise the question of beryllium neutrons at JET after installation of the ITER-like wall, and at ITER, with beryllium as the plasma facing component. This is especially important for the first few years of ITER operation, where the machine will not yet have been certified as a nuclear facility and should be run in zero-activation mode.

The magnetically confined plasmas study is one of the most promising research fields in the present days due to the high perspectives of unlimited and clean energy that fusion has generated. In this framework, the stellarator devices play a significant role because of, unlike in the tokamak case, their continuum working regime, which will be an essential feature of the future fusion commercial reactor.Heat transport studies in stellarator devices are completely necessary since the main plasma properties (and therefore, the total fusion power generated) are absolutely dependent. Nowadays, the largest stellarator in the world is the Large Helical Device (LHD). There is also a stellarator device is Spain, TJ-II, which is located in the installations of CIEMAT in Madrid. In this thesis, turbulent and conductive heat transport is studied in both devices with the aim of comparing its formation and suppression. First of all, collisional transport, i.e. neoclassical transport, which is caused by the particle collisions, is studied by means of a new transport model implemented in the transport code PRETOR-Stellarator. This model is able to calculate heat diffusivities as well as the neoclassical electric field with reasonable accuracy without spending as much computational time as in the Monte Carlo techniques. It is deduced from the results that, for both TJ-II and LHD, neoclassical transport may be quite important in plasmas with low density and high temperatures, although higher levels of neoclassical transport are obtained in TJ-II. Both devices share the feature that in low collisional plasmas, a high positive neoclassical electric field with a high shear appears in the plasma core. This electric field can be responsible of the suppression of the turbulence heat transport.Some new turbulent heat transport models have been added to PRETOR-Stellarator in order to study this kind of transport. Both, LHD and TJ-II, share a common heat transport in the confinement region (plasma core), called drift wave electromagnetic transport, and which is due to the fluctuations of the magnetic field. Outside this region, turbulent heat transport in LHD has similar characteristic to that in tokamaks, whereas in TJ-II, turbulent transport is maintained.Turbulent heat transport reduction is a major issue in fusion research, since the capability of producing commercial fusion energy depends strongly on the low levels of turbulence of the plasma. The appearance of a neoclassical electric field in the plasma core and its interaction with turbulent transport has been studied. It is shown that this electric field is able to generate a rotation in the plasma which is able to suppress turbulent transport to neoclassical levels when density is low enough. These plasmas are called to have an internal transport barrier and have stepped electron temperature profiles with hollow electron density profiles. Another important phenomenon related with electron heat transport is non-local transport, which can not be studied within the general diffusive framework that is used to study turbulent transport in plasmas. The non-local transport is caused by the interaction of long distant parts of the plasma. In this thesis, a new model for this type of transport, which is based on the convolution over a kernel of the neoclassical transport, has been proposed to explain this phenomenon. It has been shown that this model is able to simulate the main characteristics of this transport, e.g. fast pulses propagation, ballistic transport or the growing of the turbulence levels close to the axis of the device. All these phenomena have been observed in LHD and TJ-II. Finally, once electron heat transport in stellarators has been clarified, a comparison of the designs of the future commercial reactor based on both, stellarators and tokamaks, has been carried out. A stellarator commercial reactor, based on the design of the LHD, would have a 15.5 m major radius, 2.5 m minor radius, with a continuum working regime based on low temperatures and high densities. Main energy sinks are due to conductive-convective heat losses and radiation losses (in a 95% from Bremmstrahlung radiation). The fact that it has such a large major radius makes this design expensive and difficult to build. A tokamak fusion reactor would be smaller, however, the high temperatures achieved make cyclotron radiation losses to be very high, and a wall with a high reflection coefficient seems to be necessary. / L'estudi de les propietats dels plasmes confinats magnèticament esta esdevenint un dels temes primordials de recerca degut a les prometedores perspectives (de netedat i ampli abast) que l'energia produïda per fusió nuclear està fomentant. Es dins d'aquest context on l'estudi dels dispositius de confinament magnètic de tipus stellarator juguen un paper molt important, ja que un reactor de fusió basat en aquest concepte podria tenir (al contrari dels tokamaks) un mode de funcionament continu i no polsat.L'estudi del transport de calor en el dispositius de fusió per confinament magnètic de tipus stellarator és totalment necessari, ja que les propietats del plasma (i per tant de l'energia produïda per fusió) en depenen completament. Actualment, el stellarator més gran del món es troba al Japó i es diu Large Helical Device (LHD), mentre que a Espanya, el stellarator TJ-II es troba a les instal·lacions del CIEMAT a Madrid. En aquesta tesi, s'estudien ambdós dispositius per tal de determinar de que depèn que aparegui o que desaparegui el transport de calor turbulent en aquests dispositius, i si hi ha algun tipus de punt en comú.En primer lloc, s'analitza el transport de calor colisional (degut a la col·lisió de les partícules que formen el plasma) mitjançant la introducció d'un model de transport colisional (anomenat neoclàssic) al codi de transport PRETOR-Stellarator. Aquest model es capaç de calcular magnituds físiques tal com difusivitats i camps elèctrics neoclàssics però sense consumir tant de temps com a d'altres tècniques que utilitzen mètodes de Monte Carlo. Dels resultats es desprèn que el transport neoclàssic, tant al TJ-II com al LHD, pot ser important, en plasmes amb baixa densitat i temperatures grans. Ambdós dispositius comparteixen la característica de que apareix, en aquests casos, un gran camp elèctric al centre del plasma que pot ser fa que el transport turbulent disminueixi. Mitjançant la introducció de diferents models pel transport turbulent a PRETOR-Stellarator, s'estudia el transport turbulent als dos dispositius. De l'anàlisi es dedueix que ambdós dispositius poden compartir el mateix tipus de transport (anomenat electromagnètic) i que es degut a les variacions locals del camp magnètic. Tanmateix, fora de la zona central, el LHD té un tipus de transport semblant al que existeix al tokamak JET (Joint European Torus), mentre que el TJ-II continua amb el transport electromagnètic.La reducció del transport turbulent prèviament estudiat és un tema capdal ja que permetria un millor confinament del plasma. S'ha estudiat com el camp elèctric format al centre del plasma pot generar un rotació que disminueix el transport turbulent tant al LHD com al TJ-II. Aquests plasmes es diuen que tenen una barrera interna del transport que redueix el transport turbulent a valors neoclàssics sempre que la densitat sigui prou baixa.Un altre fenomen lligat al transport molt important és el transport no local, que es degut a les correlacions entre parts llunyanes del plasma i que no es pot entendre dintre del context del transport difusió que se sol emprar per a estudiar el plasmes confinats. En el marc d'aquesta tesi s'ha dissenyat un model de transport no local per mitjà d'una convolució sobre el transport neoclàssic. Amb aquest model s'ha aconseguit reproduir molts del efectes no locals que apareixen als plasmes (com ara la ràpida propagació de fenòmens turbulents o la propagació de fronts turbulents que mantenen una forma d'ona ), i que s'han descrit tant al LHD com al TJ-II.Finalment es realitza una comparació entre els dissenys dels reactors de fusió basats en stellarators i tokamaks. Un reactor de fusió stellarator tindria un radi major de 15.5 m i treballaria en mode continu amb alta densitat i baixa temperatura. Les pèrdues d'energia més importants serien, en aquest cas, degudes a la convenció i conducció dins del plasma. El fet que tingui una grandària tan gran el faria molt car de construir. En el cas dels tokamaks, la seva grandària seria més petita, però les pèrdues per radiació ciclotró serien molt grans (degut al règim d'alta temperatura i baixa densitat) i el disseny d'una paret del reactor amb un coeficient de reflexió molt gran fora totalment necessari.

Tiriamasis darbas nagrinėja fundamentalias plazmos sąveikos su termobranduolinio reaktoriaus sienelėmis problemas. Jis atliktas bendradarbiaujant su tarptautiniais tyrimų centrais Europoje, dalyvaujančiais kuriant naujos kartos termobranduolinį reaktorių. W dangos, gautos magnetroniniu garinimo būdu, buvo paveiktos argono jonų su anglies priemaišomis spinduliuote esant skirtingiems slėgiams eksperimentiniame įrenginyje. Pagrindinis dėmesys sutelktas į teigiamų jonizuotų dalelių ir plazmos sąveiką su volframo danga anglies adsorbcijos metu ir jų poveikį volframo dangos savybėms. Darbe ištirtos ir išaiškintos sąlygos, prie kurių anglis, readsorbuota iš plazmos, efektyviai pernešama į W tūrį. Darbe parodyta, kad, vykstant vienalaikei C adsorbcijai ir joninei spinduliuotei, anglis efektyviai pernešama nuo paviršiaus į tūrį, kai W paviršius tik dalinai padengtas adsorbuota anglimi. Kai W paviršius padengtas ištisine C plėvele, anglies pernešimas nuo paviršiaus į tūrį yra blokuojamas. Gautų spinduliuote paveiktų dangų savybės buvo nagrinėjamos sekančiomis technologijomis: paviršiaus topografinė analizė atlikta naudojant skenuojantį elektroninį mikroskopą, atominės jėgos mikroskopą, dangų struktūra tirta su rentgeno spindulių difraktometru, dangų profilio analizė atlikta rusenančio išlydžio optinė spektroskopija. Panaudoti analizės metodai sudarė galimybes tirti dangų elementinę sudėtį, mikrostruktūrą, paviršiaus reljefą ir atlikti elementinės sudėties profiliavimą nuo paviršiaus... [toliau žr. visą tekstą] / The dissertation was implemented in-collaboration with international research centers in Europe, who participates in the development of new century thermonuclear reactors process. The existence of more than one plasma facing material at the ITER divertor target (CFC and W) can affect substantially the fuel retention properties of W by formation of deposited hydrocarbon layers and carbides and could lead to significant changes of the mixed-material properties with respect to that of the pure W material. The task of this work is to understand these effects by means of dedicated experimental studies to determine the influence of the various physics processes in the final erosion/redeposition pattern. The goal of the work is the fabrication of W films to be used in plasma – facing components in fusion devices, and the understanding of the mechanism of physical phenomena initiating modification of mechanical properties of W – based thin films on stainless steel substrates under Ar ion irradiation and under high-flux, low-energy H+ ions irradiation in the range of temperatures. During this work there were fabricated nanocrystalline W coatings on carbon based materials (including cfcs) using magnetron sputtering technique. Main parameters: area of deposits – 20 x 17 cm, thickness homogeneity – better than 20%, coating thickness – 3-6 µm. The microstructure of w coatings was densified by deposition under continuous ion bombardment and optimization of grain size and texture to improve... [to full text]

The dissertation was implemented in-collaboration with international research centers in Europe, who participates in the development of new century thermonuclear reactors process. The existence of more than one plasma facing material at the ITER divertor target (CFC and W) can affect substantially the fuel retention properties of W by formation of deposited hydrocarbon layers and carbides and could lead to significant changes of the mixed-material properties with respect to that of the pure W material. The task of this work is to understand these effects by means of dedicated experimental studies to determine the influence of the various physics processes in the final erosion/redeposition pattern. The goal of the work is the fabrication of W films to be used in plasma – facing components in fusion devices, and the understanding of the mechanism of physical phenomena initiating modification of mechanical properties of W – based thin films on stainless steel substrates under Ar ion irradiation and under high-flux, low-energy H+ ions irradiation in the range of temperatures. During this work there were fabricated nanocrystalline W coatings on carbon based materials (including cfcs) using magnetron sputtering technique. Main parameters: area of deposits – 20 x 17 cm, thickness homogeneity – better than 20%, coating thickness – 3-6 µm. The microstructure of w coatings was densified by deposition under continuous ion bombardment and optimization of grain size and texture to improve... [to full text] / Tiriamasis darbas nagrinėja fundamentalias plazmos sąveikos su termobranduolinio reaktoriaus sienelėmis problemas. Jis atliktas bendradarbiaujant su tarptautiniais tyrimų centrais Europoje, dalyvaujančiais kuriant naujos kartos termobranduolinį reaktorių. W dangos, gautos magnetroniniu garinimo būdu, buvo paveiktos argono jonų su anglies priemaišomis spinduliuote esant skirtingiems slėgiams eksperimentiniame įrenginyje. Pagrindinis dėmesys sutelktas į teigiamų jonizuotų dalelių ir plazmos sąveiką su volframo danga anglies adsorbcijos metu ir jų poveikį volframo dangos savybėms. Darbe ištirtos ir išaiškintos sąlygos, prie kurių anglis, readsorbuota iš plazmos, efektyviai pernešama į W tūrį. Darbe parodyta, kad, vykstant vienalaikei C adsorbcijai ir joninei spinduliuotei, anglis efektyviai pernešama nuo paviršiaus į tūrį, kai W paviršius tik dalinai padengtas adsorbuota anglimi. Kai W paviršius padengtas ištisine C plėvele, anglies pernešimas nuo paviršiaus į tūrį yra blokuojamas. Gautų spinduliuote paveiktų dangų savybės buvo nagrinėjamos sekančiomis technologijomis: paviršiaus topografinė analizė atlikta naudojant skenuojantį elektroninį mikroskopą, atominės jėgos mikroskopą, dangų struktūra tirta su rentgeno spindulių difraktometru, dangų profilio analizė atlikta rusenančio išlydžio optinė spektroskopija. Panaudoti analizės metodai sudarė galimybes tirti dangų elementinę sudėtį, mikrostruktūrą, paviršiaus reljefą ir atlikti elementinės sudėties profiliavimą nuo paviršiaus... [toliau žr. visą tekstą]

Measurements of the neutron emission, resulting from nuclear fusion reactions between the hydrogen isotopes deuterium and tritium, can provide a wealth of information on the confinement properties of fusion plasmas and how these are affected by Magneto-Hydro-Dynamic (MHD) instabilities. This thesis describes work aimed to develop neutron measurement techniques for nuclear fusion plasma experiments, specifically regarding the performance and design of collimated neutron flux monitors (neutron cameras) for the Mega Ampere Spherical Tokamak, MAST, and for MAST Upgrade. The first part of the thesis focuses on the characterization of a prototype neutron camera installed at MAST and provides an account of the very first measurements of the neutron emissivity along its collimated fields of view. It is shown that the camera has sufficient temporal and spatial resolution to measure the effect of MHD instabilities on the neutron emissivity. The neutron camera fulfils the requirement on the measurements of the neutron count rate profile with less than 10 % statistical uncertainty in a time resolution of 1 ms. The instrument's more rudimentary capabilities to provide information on the neutron energy distribution are also presented and discussed. The encouraging results obtained with the prototype neutron camera show the potential of a collimated neutron flux monitor at MAST and suggest that an upgraded instrument for MAST Upgrade will provide crucial information on fast ions behavior and other relevant physics issues. The design of such an upgraded instrument for MAST Upgrade is discussed in the second part of the thesis. Two design options are explored, one consisting of two collimator arrays in the horizontal direction, another more traditional design with lines-of-sight in the poloidal cross section plane. On the basis of the experience gained with the prototype neutron camera and on the exploratory design and estimated performance for the upgraded camera presented here, a conceptual design of a neutron camera upgrade is proposed.

The promise of a vast source of energy to power the world and protect our planet using fusion technology has been the driving force for scientists and engineers around the globe for more than sixty years. Although the materialization of this ideal still in the distance, multiple scientific and technological advances have been accomplished, which have brought commercial fusion power closer to a reality than it has ever been. As part of the collaborative effort in the pursuit of realizable fusion energy, the International Thermonuclear Experimental Reactor (ITER) is being developed by a coalition of nations of which the United States is a part of. One critical technological challenge for ITER is the development of adequate plasma facing materials (PFMs) that can withstand the strenuous conditions of operation. To date, high heat flux (HHF) testing has been conducted mainly on non-irradiated specimens due to the difficulty of working with radioactive specimens, such as instrument contamination. In this thesis, the new Irradiated Material Target Station (IMTS) facility for fusion materials at Oak Ridge National Laboratory (ORNL), in which the HHFs are provided by water-wall plasma-arc lamps (PALs), is considered for neutron-irradiated specimens, especially tungsten. The facility is being used to test irradiated plasma-facing components materials for magnetic fusion reactors as part of the US-Japan plasma facing components evaluation by tritium plasma, heat and neutron irradiation experiments (PHENIX). In order to conduct HHF testing on the PFMs various sample holders designs were developed to accommodate radioactive specimens during HHF testing. As part of the effort to design sample holders that are compatible with the IMTS facility, numerical simulations were performed for different water-cooled sample holder designs with the commercial computational fluid dynamics (CFD) software package, ANSYS™ FLUENT®. The numerical models are validated against experimental temperature measurements obtained from the IMTS facility. These experimentally validated numerical models are used to assess the thermal performance of two sample holder designs and establish safe limits for HHF testing under various operating conditions. The limiting parameter for the current configuration was determined for each sample holder design. For the Gen 1 sample holder, the maximum temperature reached within the Copper rod limits the allowable incident heat flux to about 6 MW/m². In the case of the Gen 2 sample holder, the maximum temperature reached within the Molybdenum clamping disk limits the allowable incident heat flux to about 5 MW/m². In addition, the numerical model are used to parametrically investigate the effect of the operating pressure, mass flow rate, and incident heat flux on the local heat flux distributions and peak surface temperatures. Finally, a comparative analysis is conducted to evaluate the advantages and disadvantages associated with the main design modifications between the two sample holder models as to evaluate their impact in the overall thermal performance of each sample holder in order to provide conclusive recommendations for future sample holder designs.

La source d'ions des injecteurs de neutres d'ITER devra produire un fort courant d'ions négatifs de deutérium qui seront accélérés puis neutralisés afin d'obtenir un faisceau d'atomes qui chauffera le plasma thermonucléaire. Un plasma froid d'hydrogène à basse pression et forte puissance est généré par induction dans la source et les ions négatifs sont produits par des réactions en volume et en surface et extraits à travers une série de grilles électrostatiques. Cette thèse est consacrée à la modélisation de ce plasma, afin d'étudier des phénomènes peu abordés à ce jour : aspect hors équilibre des espèces neutres, déplétion et chauffage du gaz, génération et transport des atomes, et génération des ions négatifs sur les parois. Un code fluide bidimensionnel de simulation plasma a pour cela été modifié afin de simuler la cinétique des espèces neutres au moyen d'un module Direct Simulation Monte-Carlo et a été utilisé pour simuler le plasma de manière auto-cohérente. / The ion source of the ITER neutral beam injectors will have to deliver a high current of negative deuterium ions which will be accelerated and neutralized, and the resulting atom beam will heat the thermonuclear plasma. A low pressure and high power cold hydrogen plasma is inductively generated in the source and negative ions are produced by volume and surface reactions and are extracted through a set of electrostatic grids. This thesis aims at modelling this plasma, and focuses on topics that haven't been studied much before : out of equilibrium neutral kinetics, gas heating and depletion, atoms production and transport, and negative ions generation on the walls. To this end, a two-dimensional fluid plasma code has been modified to simulate the neutrals kinetics with a Direct Simulation Monte Carlo module and has been used to perform self-consistent simulations of the plasma.

The need to make maximum use of renewable resources to the detriment of fossil fuels to achieve environmental goals with an increasing energy demand is driving research into the development of technologies to obtain energy from sources that are not currently being exploited, one of them being fusion energy. The aim of this report is to provide a general overview of fusion and to provide a critical opinion on whether fusion will become a commercial energy source in the future, and if so when. The followed methodology has been a literature review complemented by an interview to B Henric M Bergsåker, teacher and researcher at the KTH on fusion plasma physics and information person for the Swedish fusion research.In the results section the fusion physics and different technological approaches have been presented. Among the studied different projects, the ITER Tokamak magnetic reactor has been selected as the most promising of these projects, as a product of international collaboration, and it has been analyzed in more detail. The obtained results have been that fusion can be an inexhaustible, environmentally friendly and safe energy source. The first-generation fusion commercial reactors are expected to be part of the energy mix before 2100.

The work focuses on Monte Carlo simulations for finding the optimal back scatter time-of-flight spectrometer design for concept testing at the NESSA facility at Uppsala University. The spectrometer consists of two scintillator detectors, D1 (placed in a neutron beam) and D2 (placed in front of D1), at some distance from each other. A fraction of the neutrons that impinge on D1 back scatter into D2 and information on the neutron energy distribution is acquired using the time-of-flight method. For the given constraints on geometry, resolution and efficiency a best resolution was found to be 6.6% with a corresponding efficiency of 1E-4 which gives a sufficient count rate for a neutron generator producing 1E+11 neutrons/s. In order to achieve a minimum of 10 000 counts/h with the same setup a D2 with an area of at least 7 cm^2 is required.