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A recent model coupling the shear-Alfv\'{e}n and acoustic continua, which depends strongly on the equilibrium shaping and on elongation in particular, is employed to explain the properties of Alfv\'{e}nic activity observed on JET plasmas below but close to the typical frequency of toroidicity-induced Alfv\'{e}n eigenmodes (TAEs). The frequency gaps predicted by the model result from high-order harmonics of the geodesic field-line curvature caused by plasma shaping (as opposed to lower-order toroidicity) and give rise to high-order geodesic acoustic eigenmodes (HOGAEs), their frequency value being close to one-half of the TAEs one. The theoretical predictions of HOGAE frequency and radial location are found to be in fair agreement with measurements in JET experiments, including magnetic, reflectometry and soft x-ray data. The stability of the observed HOGAEs is evaluated with the linear hybrid MHD/drift-kinetic code CASTOR-K, taking into account the energetic-ion populations produced by the NBI and ICRH heating systems. Wave-particle resonances, along with drive/damping mechanisms, are also discussed in order to understand the conditions leading to HOGAEs destabilization in JET plasmas., Comment: 10 pages, 11 figures, accepted for publication

Achieving nuclear fusion as an energy source on Earth is a practical goal that relies on continuing scientific and engineering innovation. Functional fusion reactors around the world today allow scientists and engineers to plan improvements that will eventually allow for greater energy output than the input required to operate the machine (including heating the plasma and operating the superconducting electromagnets that confine the plasma, among other energy inputs). The fusion reaction between nuclei of hydrogen isotopes is a carbon-free source of massive amounts of energy that could be paramount in a global turn towards greener energy. The fusion fuel needed to provide one person’s energy use for 100 years (assuming 20 kWh per day) is contained within roughly one and a half bathtubs of water and 3 laptop batteries. Given the enormous payoff of fusion, continued research and development are of great interest so that current challenges of heating and confining plasma, mitigating plasma disruptions, improving efficiency of magnets, and extending the lifetime of materials subjected to the harsh conditions surrounding the plasma may be overcome. Fusion reactor materials research carried out here at the BSC contributes to this ambitious goal. The idealistic goal for fusion materials research is to provide predictions about material behavior with the accuracy of quantum mechanical calculations at the scale of a full fusion reactor. Using strategic approximations and working at a small scale, computational fusion materials researchers can accurately reproduce and explain experimentally observed physical phenomena, such as the formation of microstructural defects in metals under neutron-irradiation, and offer the best predictions available for behavior of materials in future fusion reactor environments, where data about what will happen simply do not exist yet. In the study presented here, we examined the thermal conductivity, or how quickly a material allows heat to flow, of tungsten (W). W has been selected for plasma-facing components in ITER, which is currently under construction. We used LAMMPS atomic modelling of materials software and found that the thermal conductivity of W is significantly decreased in the presence of defects.

Nuclear fusion takes place when two light nuclei combine to make a heavier nucleus, releasing energy in the process. Magnetic confinement fusion attempts to achieve fusion and use this energy by confining the fuel in the form of a plasma. A plasma is a fully ionized gas whose behaviour is no longer dominated by short-ranged Coulomb forces, but by long-range electric and magnetic forces. Typically, plasmas are composed by hydrogen (H), helium (He), deuterium (D) or tritium (T) ions, or a combination of these. In this abstract, we tackle magnetic confinement fusion using tokamaks. Tokamaks are toroidal devices that have axial symmetry. They use poloidal and toroidal magnetic fields to create twisted magnetic field lines along which the charged particles travel in helical trajectories. In order for the plasma to reach the necessary temperatures for fusion to take place, auxiliary heating systems are used. In this abstract we focus on heating the plasma with electromagnetic waves in the ion cyclotron range of frequencies (ICRF). Significant advances have been made in the technological development of these magnetic confinement devices in order to progress towards the main goal of fusion research: to achieve electricity-producing fusion power stations that can provide energy reliably, safely and efficiently. The ratio of fusion power produced to the external power required to maintain the plasma in a steady state is known as the Q-factor. Fusion reactions provide energy to the plasma, which leads to self-heating and eventually to a self-sustained reaction, known as ignition. One of the long-term purposes of fusion research is to achieve this ignition (Q=infinite). However, no device so far has achieved a sustainable Q=1 plasma. The main candidate to do so is ITER (”The Way” in Latin), the largest tokamak nuclear fusion reactor, which is being built in the south of France and which is projected to start operating in 2025. The aim is for ITER to maintain Q≥5 and to reach Q=10 for a duration of 400-600s, demonstrating the feasibility of fusion power and of a ten-fold gain of plasma heating power. The commissioning of ITER is taking place through a staged approach. The First Plasma will be followed by an upgrade of the capabilities of the tokamak and two Pre-Fusion Power Operation (PFPO I and II) phases. In the PFPO phases, the basic controls and protection systems will be demonstrated, including the auxiliary heating and diagnostics systems, in experimental hydrogen (H) and helium (He) plasmas. The Fusion Power Operation (FPO) will start and a transition to deuterium (D) and deuterium-tritium (D-T) campaigns will be made. Most of the ICRF modelling that has been carried out so far for ITER has focused on heating scenarios relevant for the D, T and D-T plasmas in the FPO stage. There is a need to improve our understanding on the performance of ICRF in H and He plasmas in the PFPO phase of ITER, and on the heating schemes planned for this phase. In this abstract we use the ICRF heating code PION [1] integrated into the transport modelling workflow European Transport Solver (ETS) [2] to study and predict how the plasma will be heated when ICRF heating is applied to ITER PFPO plasmas. The integration into a transport modelling workflow is relevant because PION calculates the ICRF power deposition but it does not predict on its own how the heating will affect the plasma and how its parameters will evolve. A transport modelling workflow such as ETS, which has been developed inside the ITER Integrated Modelling & Analysis Suite (IMAS) [3], can calculate the evolution of the plasma discharge and provide the capabilities for self-consistent predictive simulations.

Fusion plasmas are complex systems with several physical parameters which need continuous control and adjustment. Most plasma diagnostics have limitations of applications regarding the spatial, temporal resolution with typically high relative error. A typical physical parameter is combined from the output of several diagnostics to reduce these limitations. The Atomic Beam Probe (ABP) is a novel diagnostic technique, and in our recent work, we are working on supportive numerical modelling procedures. Our tool has already been supporting diagnostic and scenario design, and we would like to give a hint about possible publication opportunities

Using the PION ICRH modelling code and comparisons against JET tokamak experiments, the effect of including pitch angle dependence within the RF diffusion operator on the fast ion particle distribution functions is quantified. It is found to be of greatest importance in cases of higher harmonic heating and lower heating ion mass, resulting in faster drop-off of the distribution's high energy tail. We see differences of several orders of magnitude in the high-energy range and significant non-linear alterations by several tens of percent to ion species power partition. ITER scenario operational parameters are also considered, and this improved treatment is shown to benefit anticipated ITER scenarios with second harmonic hydrogen heating, according to our predictions. PION's combination of benchmarked simplified wave physics and Fokker-Planck treatment offers modelling advantages. Since including the pitch angle dependence in the RF diffusion operator has not led to a significant increase in the required computing time when modelling different ICRF schemes in JET discharges, it has been made available within the production code. The CCFE part of this work has been carried out within the framework of the EUROfusion Consortium and has received funding from the Euratom research and training programme 2014–2018 and 2019–2020 under Grant Agreement No. 633053. The views and opinions expressed herein do not necessarily reflect those of the European Commission. The BSC part of this project is co-financed by the European Union Regional Development Fund within the framework of the ERDF Operational Program of Catalonia 2014–2020 with a grant of 50% of total cost eligible. The authors are grateful to Jacob Eriksson for assistance with experimental data, to Lars-Göran Eriksson for discussions on the implementation of the new features, and to Colin Roach and Michael Fitzgerald for valuable comments on the manuscript.

This paper presents a methods for the intrinsic impurity concentration measurements by means of VUV and SXR diagnostics on the JET-ILW tokamak. Measurements of mid-Z impurities content were obtained by means of VUV spectra. To provide absolute concentrations a new relative calibration technique has been proposed. It's based on cross-calibration with calibrated spectrometer by using unresolved transition array of W in the relevant wavelength range. The SXR cameras were used to deduce W profiles and poloidal asymmetries. Focus is given to hybrid discharges stopped by the real-time vessel protection system due to hot-spots formation. This effect was linked to the application of ICRH power. Local D2 gas injection allows mitigating hot-spot formation and run pulses with acceptable temperature values on vessel components. Hot spot temperature analysis showed a lower maximum temperature at higher gas rate. A decrease of impurity concentration with D2 injection rate was observed. Changes in the plasma current have a strong impact on the plasma-wall interaction, both via modifications in the edge density and in the fast-ion losses. Finally it was observed that at constant gas injection rate, both the hot spot temperature and the core impurity content decrease with the separatrix density.

This paper discusses the extension of the ‘three-ion’ species ICRF technique for heating mixture plasmas using fast injected NBI ions as resonant ‘third’ species. In this scenario the ICRF power is absorbed by the fast beam ions in the vicinity of the mode conversion layer where the left-hand polarized RF electric field E+ is strongly enhanced. The ions in the beam velocity distribution that have a Doppler-shifted resonance close to the mode conversion layer efficiently absorb RF power and undergo acceleration. We show first experimental observations of ICRF heating of D-NBI ions in H-D plasmas in JET with the ITER-like wall. In agreement with theoretical predictions and numerical modelling, acceleration of the D-NBI ions in this D-(DNBI)-H scenario is confirmed by several fast-ion measurements. An extension of the heating scheme discussed here is acceleration of T-NBI and D-NBI ions in D-T plasmas, offering the potential to further boost the Q-value in future D-T campaigns in JET. The support from the JET team is warmly acknowledged. This work has been carried out within the framework of the EUROfusion Consortium and has received funding from the Euratom research and training programme 2014-2018 under grant agreement No 633053. The views and opinions expressed herein do not necessarily reflect those of the European Commission and the ITER Organization.

A strong modelling program has been started in support of the future JET-DT campaign with the aim of guiding experiments in deuterium (D) towards maximizing fusion energy production in Deuterium–Tritium (DT). Some of the key elements have been identified by using several of the most updated and sophisticated models for predicting heat and particle transport, pedestal pressure and heating sources in an integrated modelling framework. For the high beta and low gas operational regime, the density plays a critical role and a trend towards higher fusion power is obtained at lower densities. Additionally, turbulence stabilization by E × B flow shear is shown to generate an isotope effect leading to higher confinement for DT than DD and therefore plasmas with high torque are suitable for maximizing fusion performance. Future JET campaigns will benefit from this modelling activity by defining clear priorities on their scientific program. This work has been carried out within the framework of the EUROfusion Consortium and has received funding from the Euratom research and training programme 2014–2018 under grant agreement No 633053. The views and opinions expressed herein do not necessarily reflect those of the European Commission.

We report on recent experiments on the ASDEX Upgrade (AUG) tokamak (major radius R ≈1.65 m, minor radius a ≈ 0.5 m) with third harmonic ICRF heating of deuterium beam ions. Prior to this work, the scheme has been developed and applied on the JET tokamak, the largest currently operating tokamak (R ≈ 3 m, a ≈ 1 m), for fusion product studies and for testing alpha particle diagnostics in preparation of ITER [1]. The experiments reported here demonstrate that this scheme can also be used in medium size tokamaks such as AUG despite their reduced fast ion confinement. This work has been carried out within the framework of the EUROfusion Consortium and has received funding from the Euratom research and training programme 2014-2018 under grant agreement No 633053. The views and opinions expressed herein do not necessarily reflect those of the European Commission.

This work develops strategies for adapting a particle-in-cell code to heterogeneous computer architectures and, in particular, to an ARM-based prototype of the Mont-Blanc project using OmpSs programming model and the OpenMP and OpenCL languages.

Ion cyclotron resonance frequency heating (ICRF) is one of the auxiliary heating schemes presently envisaged for ITER and DEMO. In this paper we analyse the potential of ICRF waves to heat the fuel ions in DEMO. Our analysis is carried out for the EU DEMO design¹ (B = 6.8 T, I = 18.6 MA, R = 9.25 m, a = 2.64 m) optimized for a maximum pulse length of 2.3 hrs using the ICRF modelling codes PION and TORIC [2, 3] . We focus on second harmonic heating for tritium and fundamental minority heating of ³He with a few percent of ³He in a 50%:50% D-T plasma. The dependence of the ICRF characteristics and the ICRF-accelerated ions on the ICRF and plasma parameters is investigated, giving special attention to the DEMO design point at a plasma temperature of 30 keV and an electron density of 1.2 · 10²⁰ m⁻ ³.

The JET experimental campaign has focused on studies in support of the ITER physics basis. An overview of the results obtained is given both for the reference ITER scenario, the ELMy H-mode, and for advanced scenarios which in JET are based on Internal Transport Barriers. JET studies for the ELMy H-mode have been instrumental for the definition of ITER-FEAT. Positive elongation and current scaling in the ITER scaling law have been confirmed, but the observed density scaling fits better a two term (core and edge) model. Significant progress in neo-classical tearing mode limits has been made showing that ITER operation seems to be optimised. Effective helium pumping and divertor enrichment is found to be well within ITER requirements. Target asymmetries and H-isotope retention are well simulated by modelling codes taking into account drift flows in the scrape-off plasmas. Striking improvements in fuelling effectiveness have been found with the new high field pellet launch facility. Good progress has been made on scenarios for achieving good confinement at high densities, both with RI modes and with high field side pellets. Significant development of advanced scenarios in view of their application to ITER has been achieved. Integrated advanced scenarios are in good progress with edge pressure control (impurity radiation). An access domain has been explored showing in particular that the power threshold increases with magnetic field but can be significantly reduced when Lower Hybrid current drive is used to produce target plasma with negative shear. The role of ion pressure peaking on MHD has been well documented. Lack of sufficient additional heating power and interaction with the septum at high beta prevents assessment of beta limits (steady plasmas achieved with βN up to 2.6). Plasmas with non-inductive current (INI/Ip=60%), well aligned with plasma current, high beta and good confinement have also been obtained.

Dynamics of fast minority ions energised by ion cyclotron resonance and lower hybrid heating in tokamaks is investigated by cylindrical 20 Fokker-Planck and toroidal 30 Monte Carlo calculations. The power exchange between the waves and minority ions, the ensuing fast ion bootstrap current and the power deposition are calculated in the presence of fast ion radial diffusion and thermal minority ion source near the plasma edge. Both lower hybrid and ion cyclotron heating are observed to increase by increasing overlapping of the heating regimes and by radial diffusion, because of the enhanced diffusion of the ion cyclotron heated ions from their resonance region to the peripheral lower hybrid wave region, and diffusion of thermal ions from the plasma edge to the ion cyclotron heating region. The fast minority ion current is found to be strongly modified by the bootstrap current generation in the cyclotron heating region. An analytical expression to describe the RF-induced bootstrap current is presented.

The effect of the Alfven resonance on the power deposition in fast wave current drive scenarios of ITER and possible efficiency enhancement of lower hybrid (LH) current drive by alpha particles are investigated. The mode conversion from the fast wave to the slow wave is found to remain smaller than 20%. Monte Carlo simulations show that for ITER parameters the power transfer from the alpha particles to the LH wave can be 15% of the alpha power if the LH waves can penetrate to the flux surface ρ/α = 0.5.

In recent years, a strong reduction of plasma turbulence in the presence of energetic particles has been reported in a number of magnetic confinement experiments and corresponding gyrokinetic simulations. While highly relevant to performance predictions for burning plasmas, an explanation for this primarily nonlinear effect has so far remained elusive. A thorough analysis finds that linearly marginally stable energetic particle driven modes are excited nonlinearly, depleting the energy content of the turbulence and acting as an additional catalyst for energy transfer to zonal modes (the dominant turbulence saturation channel). Respective signatures are found in a number of simulations for different JET and ASDEX Upgrade discharges with reduced transport levels attributed to energetic ion effects. The simulations presented in this work were performed at the Cobra HPC system at the Max Planck Computing and Data Facility (MPCDF), Germany. This work has been carried out within the framework of the EUROfusion Consortium and has received funding from the Euratom research and training programme 2014-2018 and 2019-2020 under grant agreement No. 633053. The views and opinions expressed herein do not necessarily reflect those of the European Commission. Furthermore, we acknowledge the CINECA award under the ISCRA initiative, for the availability of high performance computing resources and support. The authors would like to thank N. Bonanomi, J. Citrin, H. Doerk, Ph. Lauber, P. Manas, P. Mantica, M.J. Pueschel, K. Stimmel, P.W. Terry, D. Zarzoso, and A. Zocco for all the stimulating discussions, useful suggestions and comments.

One of the challenges that magnetic confinement fusion faces is achieving extreme temperatures inside the reactors. Absorption of electromagnetic waves in the ion cyclotron range of frequencies (ICRF) has demonstrated efficient plasma heating in present-day tokamak experiments and it is one of the three auxiliary heating methods foreseen for the International Thermonuclear Experimental Reactor (ITER). Therefore, the study of different heating schemes using ICRF waves is of utmost interest to optimize the fusion performance. The present thesis is focused in the computational assessment of plasma heating using ICRF waves and neutral beam injection (NBI) using the PION code. A strong emphasis is given to the physics behind these heating mechanisms and how they affect the fusion performance. This project has been carried out in the context of the present deuterium (D) campaign that is being performed at the Joint European Torus (JET) experimental reactor in preparation for the next deuterium-tritium (D-T) campaign DTE2, which is planned to begin in 2021. The results presented in this thesis consist of two parts. In the first part we model several D plasma discharges carried out at JET and we assess the role of different heating characteristics in the fusion yield in pure D plasmas. The second part of this project is focused on the extrapolation of a high performance plasma discharge to a 50%:50% D-T scenario. A comparison of the heating characteristics with pure D plasmas is provided, and special attention is given to the resulting fusion yield. Uno de los retos que afronta la fusión por confinamiento magnético es conseguir temperaturas extremas dentro de los reactores. La absorción de ondas electromagnéticas en el rango de frecuencias de ciclotrón de los iones (ICRF) ha demostrado ser un método eficiente para calentar el plasma en experimentos actuales de tokamaks. Por lo tanto, el estudio de diferentes esquemas de calentamiento con ondas ICRF es de especial relevancia para optimizar el rendimiento de la fusión. Este trabajo se centra en el análisis computacional del calentamiento del plasma con ondas ICRF y neutral beam injection (NBI) mediante el código PION. Se ha dado énfasis en la física detrás los mecanismos de calentamiento y cómo afectan al rendimiento de la fusión. Éste proyecto se ha llevado a cabo en el contexto de la presente campaña de plasmas de deuterio (D) en el reactor experimental Joint European Torus (JET) en preparación para la siguiente campaña de plasmas de deuterio-tritio (D-T), DTE2, prevista para 2021. Los resultados de éste trabajo se han presentado en dos partes. En la primera, se modelan varias descargas de plasma de D que han tenido lugar en JET y analizamos el rol del calentamiento en la potencia obtenida por fusión en plasmas de D. La segunda parte del proyecto consiste en la extrapolación de una descarga de alto rendimiento a un escenario de 50%:50% D-T, donde se compara el calentamiento con los plasmas de D y se analiza la potencia equivalente obtenida por fusión. Un dels reptes que afronta la fusió per confinament magnètic és aconseguir temperatures extremes dins dels reactors. L'absorció d'ones electromagnètiques en el rang de freqüències de ciclotró dels ions (ICRF) ha demostrat ser un mètode eficient per escalfar el plasma en experiments actuals de tokamaks. Per tant, l'estudi de diferents esquemes d'escalfament amb ones ICRF és d'especial rellevància per optimitzar el rendiment de la fusió. Aquest treball se centra en l'anàlisi computacional de l'escalfament del plasma amb ones ICRF i neutral beam injection (NBI) mitjançant el codi PION. S'ha donat èmfasi en la física darrere els mecanismes d'escalfament i com afecten el rendiment de la fusió. Aquest projecte s'ha dut a terme en el context de la present campanya de plasmes de deuteri (D) en el reactor experimental Joint European Torus (JET) en preparació per a la següent campanya de plasmes de deuteri-triti (D-T), DTE2, prevista per 2021. els resultats d'aquest treball s'han presentat en dues parts. En la primera, es modelen diverses descàrregues de plasma de D que han tingut lloc en JET i analitzem el paper de l'escalfament en la potència obtinguda per fusió en plasmes de D. La segona part de el projecte consisteix en l'extrapolació d'una descàrrega d'alt rendiment a un escenari de 50%: 50% D-T, on es compara l'escalfament amb els plasmes de D i s'analitza la potència equivalent obtinguda per fusió.

Tesis sotmesa a embargament des de la defensa fins al dia 1 de setembre de 2020 Heating plasmas to a relevant fusion temperature is one of the key aspects of magnetically confined fusion plasmas. Radio frequency (RF) heating with electromagnetic waves in the ion cyclotron range of frequencies (ICRF) has been proven to be an efficient auxiliary method in present fusion devices such as tokamaks. Moreover, the International Thermonuclear Experimental Reactor (ITER) will be provided with ICRF antennas as one of the main heating mechanisms. For that, the study of different heating schemes to optimise the fusion performance is of utmost importance. During the 2015-2019 Joint European Torus (JET) campaigns many efforts have been devoted to the exploration of high-performance plasma scenarios envisaged for D-T operation in JET. Experiments in D, H and T are expected to lead in 2020 to the first experiments with 50%:50% D-T mixtures. These last campaigns at JET have been focused on enhancing the fusion performance of the baseline and hybrid scenarios with the final goal of improving ITER's future operation. This thesis reports on the modelling study of plasma heating through ICRF waves and NBI for recent experiments at JET with special emphasis on plasma performance. The modelling has been performed mainly with the ICRF code PION. Simulations are in excellent agreement with experimental results which proves the reliability of the results shown in this thesis. The assessment of the results offer an overview to understand and optimise plasma performance for high-performance hybrid discharges that were performed with D plasma and H minority. Impurity accumulation control with ICRF waves was found to be efficient only for a range of central resonance locations while impurity accumulation occurred for off-axis resonance. Contribution to temperature screening from fast ions was calculated to be negligible when finite orbit widths (FOW) are taken into account, as opposed to previous studies that did not take into account FOW. Small differences in H concentration have a large impact on power partition between H and D. The lower the H concentration the larger the power channeled to D which is shown to substantially enhance the D-D fusion rate. The study of a neutron record high-performance discharge shows high bulk ion heating and low H concentration as key ingredients for increased fusion performance. Of especial relevance for ITER is the study of the D-T prediction from high-performance discharges. This analysis compares two ICRF schemes, H and 3He minority. It is shown that 3He is a strong absorber and provides higher bulk ion heating as compared to H. However, ICRF fusion enhancement is computed to be larger in H, as this scheme has a stronger 2nd harmonic heating. In D-T, ICRF fusion enhancement is computed to be significantly lower with regards to D-D plasmas due to different fusion cross sections. Results in preparation of the T and D-T campaigns at JET show that the extrapolation from T to D-T plasmas is not straightforward. PION predicts the T density to have a large impact on the T velocity distribution function for the ITER relevant 2nd$ T harmonic heating scheme. Larger concentrations of T lead to higher bulk ion heating, therefore, it is expected D-T bulk ion heating to be lower. L'escalfament de plasmes a una temperatura de fusió rellevant és un dels aspectes clau dels plasmes de fusió confinats magnèticament. L'escalfament de radiofreqüència (RF) amb ones electromagnètiques en el rang de freqüències iòniques (ICRF) s'ha demostrat que és un mètode auxiliar eficient en dispositius de fusió actuals, com ara tokamaks. D'altra banda, el Reactor Termonuclear Experimental Internacional (ITER) disposarà de les antenes ICRF com un dels principals mecanismes d'escalfament. Per això, l'estudi de diferents esquemes d'escalfament per optimitzar el rendiment de fusió és sumament important. Durant les campanyes del Joint European Torus (JET) 2015-2019 s'han dedicat molts esforços a l'exploració d'escenaris de plasma d'alt rendiment previstos per a l'operació D-T en JET. S'espera que els experiments en D, H i T condueixin el 2020 als primers experiments amb 50%: 50% de mescles D-T. Aquestes darreres campanyes a JET s'han centrat en millorar el rendiment de fusió dels escenaris bàsics i híbrids amb l'objectiu final de millorar el futur funcionament de l'ITER. Aquesta tesi dóna informació sobre l'estudi de modelització de l'escalfament de plasma mitjançant ones ICRF i NBI per a experiments recents a JET amb especial èmfasi en el rendiment del plasma. El modelat s'ha realitzat principalment amb el codi PION per a ICRF. Les simulacions estan en excel·lent acord amb els resultats experimentals el qual demostra la fiabilitat dels resultats mostrats en aquesta tesi. L'avaluació dels resultats ofereix una visió general per comprendre i optimitzar el rendiment del plasma per a descàrregues híbrides d'alt rendiment que es van realitzar amb plasma D i una minoria d'H. El control de l'acumulació de les impureses amb ones ICRF només es va trobar eficaç per a una sèrie de localitzacions de ressonància central, mentre que es va produir una acumulació d'impureses per a ressonàncies fora de l'eix. Es va calcular que la contribució a l'apantallament per temperatura de les impureses pels ions ràpids és insignificant quan es tenen en compte les amplades d'òrbita finites (FOW), a diferència d'estudis anteriors que no tenien en compte FOW. Petites diferències de concentració d'H tenen un gran impacte en la partició de potència entre H i D. Com més baixa sigui la concentració d'H, més gran és la potència canalitzada a D, es demostra que millora substancialment el número de reaccions de fusió D-D. L'estudi d'una descàrrega d'alt rendiment amb un rècord de neutrons generats mostra que l'escalfament dels ions i una baixa concentració d'H són els ingredients clau per augmentar el rendiment de les reaccions de fusió. És especialment rellevant per a ITER l'estudi de la predicció de D-T de descàrregues d'alt rendiment. Aquesta anàlisi compara dos esquemes d'ICRF, la minoria d'H i la de 3He. Està demostrat que 3He és un gran absorbent i proporciona un major escalfament dels ions en comparació amb l'H. No obstant això, la millora de la fusió per ICRF és més gran en H, ja que aquest esquema té un segon escalfament harmònic més potent. A D-T, la millora de la fusió per l'ICRF és significativament menor pel que fa als plasmes D-D a causa de les diferències en les seccions eficaces de les reaccions de fusió. Els resultats de la preparació de les campanyes de T i D-T a JET demostren que l'extrapolació de plasmes de T a D-T no és senzilla. PION pronostica que la densitat de T tindrà un gran impacte en la funció de distribució de velocitat del T per a l'esquema d'escalfament harmònic pertanyent a ITER. Per exemple, les concentracions més grans de T generen un escalfament de ions més gran, per tant, es preveu que l'escalfament dels ions a D-T sigui més baix.

We introduce a new deterministic neutron transport code integrated in the Alya system at BSC. We explain the basic physical concepts responsible for the production of neutrons and derive the strong and weak forms of the Neutron Transport Equation along with some simplifications. We also make a brief introduction to the codes used by the author. Two numerical calculations with our software and other tools are performed. We also check the performance of the code. With this work we provide a good starting point for the further development of the Alya system for multiphysics applications in the fusion domain. Presentem un nou codi determinista de transport de neutrons integrat dins del sistema Alya del BSC. S'expliquen els principis físics bàsics de producció de neutrons i es dedueixen les versions forta i dèbil de l'equació de transport de neutrons juntament amb algunes simplificacions. Presentem dues simulacions comparades entre el nostres software i d'altres. També estudiem el rendiment del codi. Aquesta tesi suposa un bon punt de partida per a un futur desenvolupament del sistema Alya per a problemes de multifísica dins del camp de la fusió nuclear.

Ion cyclotron resonance frequency heating (ICRH) is one of the most important mechanisms to heat fusion plasma. The magnetic field generated by the magnetic coils forces ions to follow a cyclotron trajectory around the magnetic field lines due to the Lorentz force. Therefore, ions revolve around the magnetic field lines in a determined frequency, the so called ion cyclotron frequency. ICRH is based on launching electromagnetic waves from the low-field side in such a way that their frequency matches the one from ions cyclotron frequency. When both frequencies match, another effect begins to occur, the wave-particle interaction. At this point, ions start damping the wave by absorbing its energy. This effect modifies the distribution function of ions which develops a tail in the high energy region. The fast ions produced by the energy absorption from the electromagnetic waves play an important role in heating the bulk plasma. Therefore, it is crucial to know how the energy of the wave is distributed among ions and electrons, and how the fast ions produced deliver their energy to the other particles, ions and electrons. This Msc thesis is a first computational assessment of bulk plasma heating for DEMO. The DEMOnstration power plant is a proposed nuclear fusion power plant that is expected to be built after the experimental reactor ITER. It will be the first fusion reactor to produce electrical energy. Its parameters and scope are still not fixed yet, a few different yet similar designs exist. However, the physical dimensions and energy output in DEMO are much bigger than that of ITER. In fact, DEMO’s 2 to 4 gigawatts of fusion power will be in the scale of the modern electric power plants. In this sense, the analysis here presented, takes into account the evolution of the fast ions and assesses their behavior at DEMO. The ICRH scenarios studied are the second harmonic tritium with and without 3He in D-T plasma as they are regarded as the most promising ICRH scenarios. Plasma parameters, as temperature T and electron density ne, are scanned in order to obtain the behavior of the fast ions. A DEMO design point which has been used to perform deeper analysis for a determined case has been established at T = 30 keV , ne = 1.2 · 1020 m−3. The analysis has been carried out using the PION code. PION has been extensively benchmarked against JET results. It is able to solve the evolution in time of the distribution function and to compute the absorption of the electromagnetic wave. This problem must be solved self-consistently. PION simulations consist in a number of time steps. First of all, for each time step, the power absorbed is calculated. This information is then used for computing the distribution function with a Fokker-Planck model, which will be used to compute the absorption power at the beginning of the next time step. This process is repeated iteratively until convergence is reached. The results obtained in this work are two: i) the bulk ion heating and ii) fast ion parameters. We have noticed that by placing the resonance region slightly closer to the outer side the bulk ion heating is improved in both scenarios by reducing the direct electron damping. The values for bulk ion heating of a 100 MW electromagnetic wave launch are 55.84 MW for a plasma with 3% of 3He and 43.00 MW for a plasma without 3He. The plasma without the 3He dilution shows a higher reaction rate and also its fast ions are considerably more energetic. So, as the minority heating scenario has an enhanced bulk ion heating, the second harmonic tritium scenario presents two advantages, firstly that no 3He is required and secondly that there is no 3He dilution.