Academic literature on the topic 'Plasmi in tokamak'
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Journal articles on the topic "Plasmi in tokamak"
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Liang, Chen, Zhuang Ma, Zhen Sun, Xiaoman Zhang, Xin You, Zhuang Liu, Guizhong Zuo, Jiansheng Hu, and Yan Feng. "Demonstration of object location, classification, and characterization by developed deep learning dust ablation trail analysis code package using plasma jets." Review of Scientific Instruments 94, no. 2 (February 1, 2023): 023506. http://dx.doi.org/10.1063/5.0123614.
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Based on deep learning, a Dust Ablation Trail Analysis (DATA) code package is developed to detect dust ablation trails in tokamaks, which is intended to analyze a large amount data of tokamak dusts. To validate and benchmark the DATA code package, 2440 plasma jet images are exploited for the training and test of the deep learning DATA code package, since plasma jets resemble the shape and size of dust ablation clouds in tokamaks. After being trained by 1920 plasma jet images, the DATA code package is able to locate 100% plasma jets, classify plasma jets with the accuracy of >99.9%, and output image skeleton information for classified plasma jets. The DATA code package trained by the plasma jet images is also used to analyze the dust ablation trails captured in the Experimental Advanced Superconducting (EAST) tokamak with the satisfactory performance, further verifying its applicability in the fusion dust ablation investigation. Based on its excellent performance presented here, it is demonstrated that our DATA code package is able to automatically identify and analyze dust ablation trails in tokamaks, which can be used for further detailed investigations, such as the three-dimensional reconstruction of dusts and their ablation trails.
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Pankratov, Igor M., and Volodymyr Y. Bochko. "Nonlinear Cone Model for Investigation of Runaway Electron Synchrotron Radiation Spot Shape." 3, no. 3 (September 28, 2021): 18–24. http://dx.doi.org/10.26565/2312-4334-2021-3-02.
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The runaway electron event is the fundamental physical phenomenon and tokamak is the most advanced conception of the plasma magnetic confinement. The energy of disruption generated runaway electrons can reach as high as tens of mega-electron-volt and they can cause a catastrophic damage of plasma-facing-component surfaces in large tokamaks and International Thermonuclear Experimental Reactor (ITER). Due to its importance, this phenomenon is being actively studied both theoretically and experimentally in leading thermonuclear fusion centers. Thus, effective monitoring of the runaway electrons is an important task. The synchrotron radiation diagnostic allows direct observation of such runaway electrons and an analysis of their parameters and promotes the safety operation of present-day large tokamaks and future ITER. In 1990 such diagnostic had demonstrated its effectiveness on the TEXTOR (Tokamak Experiment for Technology Oriented Research, Germany) tokamak for investigation of runaway electrons beam size, position, number, and maximum energy. Now this diagnostic is installed practically on all the present-day’s tokamaks. The parameter v┴/|v||| strongly influences on the runaway electron synchrotron radiation behavior (v|| is the longitudinal velocity, v┴ is the transverse velocity with respect to the magnetic field B). The paper is devoted to the theoretical investigation of runaway electron synchrotron radiation spot shape when this parameter is not small that corresponds to present-day tokamak experiments. The features of the relativistic electron motion in a tokamak are taken into account. The influence of the detector position on runaway electron synchrotron radiation data is discussed. Analysis carried out in the frame of the nonlinear cone model. In this model, the ultrarelativistic electrons emit radiation in the direction of their velocity v→ and the velocity vector runs along the surface of a cone whose axis is parallel to the magnetic field B. The case of the small parameter v┴/|v||| (v┴/|v|||<<1, linear cone model) was considered in the paper: Plasma Phys. Rep. 22, 535 (1996) and these theoretical results are used for experimental data analysis.
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Windridge, Melanie. "Smaller and quicker with spherical tokamaks and high-temperature superconductors." Philosophical Transactions of the Royal Society A: Mathematical, Physical and Engineering Sciences 377, no. 2141 (February 4, 2019): 20170438. http://dx.doi.org/10.1098/rsta.2017.0438.
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Research in the 1970s and 1980s by Sykes, Peng, Jassby and others showed the theoretical advantage of the spherical tokamak (ST) shape. Experiments on START and MAST at Culham throughout the 1990s and 2000s, alongside other international STs like NSTX at the Princeton Plasma Physics Laboratory, confirmed their increased efficiency (namely operation at higher beta) and tested the plasma physics in new regimes. However, while interesting devices for study, the perceived technological difficulties due to the compact shape initially prevented STs being seriously considered as viable power plants. Then, in the 2010s, high-temperature superconductor (HTS) materials became available as a reliable engineering material, fabricated into long tapes suitable for winding into magnets. Realizing the advantages of this material and its possibilities for fusion, Tokamak Energy proposed a new ST path to fusion power and began working on demonstrating the viability of HTS for fusion magnets. The company is now operating a compact tokamak with copper magnets, R 0 ∼ 0.4 m, R / a ∼ 1.8, and target I p = 2MA, B t0 = 3 T, while in parallel developing a 5 T HTS demonstrator tokamak magnet. Here we discuss why HTS can be a game-changer for tokamak fusion. We outline Tokamak Energy's solution for a faster way to fusion and discuss plans and progress, including benefits of smaller devices on the development path and advantages of modularity in power plants. We will indicate some of the key research areas in compact tokamaks and introduce the physics considerations behind the ST approach, to be further developed in the subsequent paper by Alan Costley. This article is part of a discussion meeting issue ‘Fusion energy using tokamaks: can development be accelerated?’.
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Garrido, I., A. J. Garrido, M. G. Sevillano, and J. A. Romero. "Robust Sliding Mode Control for Tokamaks." Mathematical Problems in Engineering 2012 (2012): 1–14. http://dx.doi.org/10.1155/2012/341405.
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Nuclear fusion has arisen as an alternative energy to avoid carbon dioxide emissions, being the tokamak a promising nuclear fusion reactor that uses a magnetic field to confine plasma in the shape of a torus. However, different kinds of magnetohydrodynamic instabilities may affect tokamak plasma equilibrium, causing severe reduction of particle confinement and leading to plasma disruptions. In this sense, numerous efforts and resources have been devoted to seeking solutions for the different plasma control problems so as to avoid energy confinement time decrements in these devices. In particular, since the growth rate of the vertical instability increases with the internal inductance, lowering the internal inductance is a fundamental issue to address for the elongated plasmas employed within the advanced tokamaks currently under development. In this sense, this paper introduces a lumped parameter numerical model of the tokamak in order to design a novel robust sliding mode controller for the internal inductance using the transformer primary coil as actuator.
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Saperstein, A. R., J. P. Levesque, M. E. Mauel, and G. A. Navratil. "Halo current rotation scaling in post-disruption plasmas." Nuclear Fusion 62, no. 2 (January 6, 2022): 026044. http://dx.doi.org/10.1088/1741-4326/ac4186.
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Abstract Halo current (HC) rotation during disruptions can be potentially dangerous if resonant with the structures surrounding a tokamak plasma. We propose a drift-frequency-based scaling law for the rotation frequency of the asymmetric component of the HC as a function of toroidal field strength and plasma minor radius (f rot ∝ 1/B T a 2). This scaling law is consistent with results reported for many tokamaks and is motivated by the faster HC rotation observed in the HBT-EP tokamak. Projection of the rotation frequency to ITER and SPARC parameters suggest the asymmetric HC rotation will be on the order of 10 Hz and 60 Hz, respectively.
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Garrido, Izaskun, Aitor J. Garrido, Jesús A. Romero, Edorta Carrascal, Goretti Sevillano-Berasategui, and Oscar Barambones. "Low EffortLiNuclear Fusion Plasma Control Using Model Predictive Control Laws." Mathematical Problems in Engineering 2015 (2015): 1–8. http://dx.doi.org/10.1155/2015/527420.
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One of the main problems of fusion energy is to achieve longer pulse duration by avoiding the premature reaction decay due to plasma instabilities. The control of the plasma inductance arises as an essential tool for the successful operation of tokamak fusion reactors in order to overcome stability issues as well as the new challenges specific to advanced scenarios operation. In this sense, given that advanced tokamaks will suffer from limited power available from noninductive current drive actuators, the transformer primary coil could assist in reducing the power requirements of the noninductive current drive sources needed for current profile control. Therefore, tokamak operation may benefit from advanced control laws beyond the traditionally used PID schemes by reducing instabilities while guaranteeing the tokamak integrity. In this paper, a novel model predictive control (MPC) scheme has been developed and successfully employed to optimize both current and internal inductance of the plasma, which influences the L-H transition timing, the density peaking, and pedestal pressure. Results show that the internal inductance and current profiles can be adequately controlled while maintaining the minimal control action required in tokamak operation.
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Podpaly, Y. A., J. E. Rice, P. Beiersdorfer, M. L. Reinke, J. Clementson, and H. S. Barnard. "Tungsten measurement on Alcator C-Mod and EBIT for future fusion reactors1This article is part of a Special Issue on the 10th International Colloquium on Atomic Spectra and Oscillator Strengths for Astrophysical and Laboratory Plasmas." Canadian Journal of Physics 89, no. 5 (May 2011): 591–97. http://dx.doi.org/10.1139/p11-038.
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Tungsten will be an important element in nearly all future fusion reactors because of its presence in plasma facing components. This makes tungsten a good candidate for a diagnostic element for ion temperature and toroidal velocity measurement, and it makes understanding tungsten emissions important for tokamak power balance. The effect of tungsten on tokamak plasmas is investigated at the Alcator C-Mod tokamak using VUV, bolometry, and soft X-ray spectroscopy. Tungsten was present in Alcator C-Mod as a plasma facing component and through laser blow-off impurity injection. Quasi-continuum emission previously seen at other tokamaks has been identified. Theoretical predictions are presented of tungsten emission that could be expected in future Alcator C-Mod measurements. Furthermore, spectra of highly charged tungsten ions have been studied at the SuperEBIT electron beam ion trap. This emission could prove useful for spectroscopic diagnostics of future high-temperature fusion reactor plasmas.
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Federici, Fabio, Matthew L. Reinke, Bruce Lipschultz, Andrew J. Thornton, James R. Harrison, Jack J. Lovell, and Matthias Bernert. "Design and implementation of a prototype infrared video bolometer (IRVB) in MAST Upgrade." Review of Scientific Instruments 94, no. 3 (March 1, 2023): 033502. http://dx.doi.org/10.1063/5.0128768.
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A prototype infrared video bolometer (IRVB) was successfully deployed in the Mega Ampere Spherical Tokamak Upgrade (MAST Upgrade or MAST-U), the first deployment of such a diagnostic in a spherical tokamak. The IRVB was designed to study the radiation around the lower x-point, another first in tokamaks, and has the potential to estimate emissivity profiles with spatial resolution beyond what is achievable with resistive bolometry. The system was fully characterized prior to installation on MAST-U, and the results are summarized here. After installation, it was verified that the actual measurement geometry in the tokamak qualitatively matches the design; this is a particularly difficult process for bolometers and was done using specific features of the plasma itself. The installed IRVB measurements are consistent both with observations from other diagnostics, including magnetic reconstruction, visible light cameras, and resistive bolometry, as well as with the IRVB-designed view. Early results show that with conventional divertor geometry and only intrinsic impurities (for example, C and He), the progression of radiative detachment follows a similar path to that observed for large aspect ratio tokamaks: The peak of the radiation moves along the separatrix from the targets to the x-point and high-field side midplane with a toroidally symmetric structure that can eventually lead to strong effects on the core plasma inside the separatrix.
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Dlougach, Eugenia, Alexander Panasenkov, Boris Kuteev, and Arkady Serikov. "Neutral Beam Coupling with Plasma in a Compact Fusion Neutron Source." Applied Sciences 12, no. 17 (August 23, 2022): 8404. http://dx.doi.org/10.3390/app12178404.
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FNS-ST is a fusion neutron source project based on a spherical tokamak (R/a = 0.5 m/0.3 m) with a steady-state neutron generation of ~1018 n/s. Neutral beam injection (NBI) is supposed to maintain steady-state operation, non-inductive current drive and neutron production in FNS-ST plasma. In a low aspect ratio device, the toroidal magnetic field shape is not optimal for fast ions confinement in plasma, and the toroidal effects are more pronounced compared to the conventional tokamak design (with R/a > 2.5). The neutral beam production and the tokamak plasma response to NBI were efficiently modeled by a specialized beam-plasma software package BTR-BTOR, which allowed fast optimization of the neutral beam transport and evolution within the injector unit, as well as the parametric study of NBI induced effects in plasma. The “Lite neutral beam model” (LNB) implements a statistical beam description in 6-dimensional phase space (106–1010 particles), and the beam particle conversions are organized as a data flow pipeline. This parametric study of FNS-ST tokamak is focused on the beam-plasma coupling issue. The main result of the study is a method to achieve steady-state current drive and fusion controllability in beam-driven toroidal plasmas. LNB methods can be also applied to NBI design for conventional tokamaks.
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Mitrishkin, Yuri V., Valerii I. Kruzhkov, and Pavel S. Korenev. "Methodology of Plasma Shape Reachability Area Estimation in D-Shaped Tokamaks." Mathematics 10, no. 23 (December 5, 2022): 4605. http://dx.doi.org/10.3390/math10234605.
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This paper suggests and develops a new methodology of estimation for a multivariable reachability region of a plasma separatrix shape on the divertor phase of a plasma discharge in D-shaped tokamaks. The methodology is applied to a spherical Globus-M/M2 tokamak, including the estimation of a controllability region of a vertical unstable plasma position on the basis of the experimental data. An assessment of the controllability region and the reachability region of the plasma is important for the design of tokamak poloidal field coils and the synthesis of a plasma magnetic control system. When designing a D-shaped tokamak, it is necessary to avoid the small controllability region of the vertically unstable plasma, because such cases occur in practice at a restricted voltage on a horizon field coil. To make the estimations mentioned above robust, PID-controllers for vertical and horizontal plasma position control were designed using the Quantitative Feedback Theory approach, which stabilizes the system and provides satisfactory control indexes (stability margins, setting time, overshoot) during plasma discharges. The controllers were tested on a series of plasma models and nonlinear models of current inverters in auto-oscillation mode as actuators for plasma position control. The estimations were made on these models, taking into account limitations on control actions, i.e., voltages on poloidal field coils. This research is the first step in the design of the plasma shape feedback control system for the operation of the Globus-M2 spherical tokamak. The developed methodology may be used in the design of poloidal field coil systems in tokamak projects in order to avoid weak achievability and controllability regions in magnetic plasma control. It was found that there is a strong cross-influence from the PF-coils currents and the CC current on the plasma shape; hence, these coils should be used to control the plasma shape simultaneously.
Dissertations / Theses on the topic "Plasmi in tokamak"
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CASIRAGHI, IRENE. "First principle based integrated modelling in support of the Divertor Tokamak Test facility design." Doctoral thesis, Università degli Studi di Milano-Bicocca, 2023. https://hdl.handle.net/10281/402360.
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Nel programma di ricerca europeo per la fusione termonucleare controllata sono stati definiti otto differenti obiettivi a lungo termine. Una di queste sfide cruciali riguarda lo smaltimento (exhaust) di particelle ed energia provenienti da un reattore a fusione.
Per sviluppare e testare delle strategie alternative atte a risolvere il problema dell'exhaust, una nuova macchina sperimentale è attualmente in costruzione in Italia a Frascati presso il centro di ricerca ENEA: il Divertor Tokamak Test facility (DTT).
Per progettare un nuovo tokamak sono richiesti sforzi congiunti di fisici ed ingegneri. Al fine di ridurre i costi e minimizzare i rischi, uno strumento essenziale è la modellizzazione integrata il più completa possibile basata su principi primi.
Il presente progetto di dottorato è incentrato sullo sviluppo di simulazioni multi-canale basate sulla fisica dei principali scenari operazionali di riferimento di DTT.
Modelli all'avanguardia di trasporto, riscaldamento, fuelling ed equilibrio magnetico vengono integrati in queste simulazioni per predire in modo auto-consistente profili di plasma e parametri di scenario. Vengono anche calcolate tutte le interazioni non lineari tra sistemi di riscaldamento e plasma e tra i diversi canali di trasporto.
Durante questo lavoro, le simulazioni di DTT sono state progressivamente migliorate perfezionandone le impostazioni e includendo un crescente numero di aspetti grazie all'aggiunta di codici appositi. Inoltre sono stati inclusi man mano aggiornamenti dei sistemi di riscaldamento, dell'equilibrio magnetico e della configurazione della macchina per seguire l'evoluzione del progetto.
Il confronto tra simulazioni analoghe con differenti modelli quasi-lineari di trasporto ci rende fiduciosi dell'affidabilità dei profili di plasma predetti e ci permette di identificare i punti deboli dei modelli nei vari regimi in cui opera DTT. Questi modelli quasi-lineari sono stati inoltre validati mediante simulazioni girocinetiche nel range di parametri di DTT.
L'accuratezza delle predizioni è state migliorata in modo ricorsivo accordando le condizioni al contorno delle simulazioni di core e delle simulazioni del SOL, garantendo così una consistenza core-edge-SOL. Abbiamo studiato lo scenario a massime performance per guidare la progettazione della macchina e il primo plasma e gli scenari intermedi per assistere le fasi iniziali.
Le performance dello scenario a piena potenza è stato testato con nove differenti opzioni di riscaldamento allo scopo di selezionare la distribuzione di potenza ottimale tra i tre sistemi di riscaldamento ausiliario. È stata poi verificata la compatibilità dello scenario a piena potenza con le capacità del sistema di bobine elettromagnetiche. Inoltre per la prima volta sono stati stimati, nello scenario a massima potenza, i denti di sega e gli ELMs di DTT. Un'analisi delle prestazioni richieste ai sistemi di fuelling per sostenere gli alti profili di densità ha dimostrato che sarebbe insufficiente utilizzare solamente un sistema di gas puffing e che sono necessari pellet di deuterio per alimentare DTT. Sono stati stimati i tassi di emissione neutronica, risultando compatibili con il progetto attuale delle schermature neutroniche.
Questo progetto di dottorato ha portato all'ottimizzazione delle dimensioni della macchina e alla definizione delle potenze di riferimento dei sistemi di riscaldamento e ha fornito i profili di riferimento per la progettazione delle diagnostiche, la stima delle rese neutroniche, il calcolo delle perdite di particelle veloci, i requisiti del gas puffing e/o dei pellet per il fuelling, valutazioni MHD e altri lavori.The European research roadmap towards thermonuclear fusion energy defined eight different missions to guide the long–term programme. One of these crucial challenges is the controlled power and particle exhaust from a fusion reactor. To develop and test alternative strategies to solve the exhaust problem, in Italy a new experimental device is now under construction at the ENEA Research Center in Frascati: the Divertor Tokamak Test facility (DTT). Designing a new tokamak requires concerted efforts of physicists and engineers. To reduce costs and minimise risks, a first–principle based integrated modelling as comprehensive as possible of plasma discharges is an essential tool. The focus of this PhD project was to perform the first physics–based multi–channel simulations of the main baseline operational scenarios of DTT. In these simulations state–of–art modules for transport, heating, fuelling, and magnetic equilibrium are integrated to achieve self–consistent predictions of plasma profiles and scenario parameters. All non–linear interactions between heating and plasma and between the different transport channels are also calculated. During this work, the DTT simulations have been progressively enhanced adding codes to include a growing number of aspects and refining run settings. Moreover, updates of the heating systems, magnetic equilibria, and device configuration have been included to comply with the evolving machine design. The comparison among analogous simulations with different quasi–linear transport models made us confident in the reliability of the predicted plasma profiles and allowed us to identify the weak points of the models in the various DTT operational regimes. A validation of these quasi–linear models against the gyrokinetic simulations in the specific DTT range of parameters was also performed. The prediction accuracy has been improved recursively by matching the core and SOL simulation boundary conditions to guarantee the core–edge–SOL consistency. We investigated the full performance scenario to guide the machine design, and the first plasma and intermediate scenarios to assist the commissioning phases. The full performance scenario was tested with nine different heating mix options to select the optimal power distribution amongst the three auxiliary heating systems. The compatibility of the full power scenario with the electromagnetic coil system capabilities was then verified. In addition, the DTT sawteeth and ELMs during the full power scenario were estimated for the first time. An analysis of the required fuelling system performance to sustain the high density profiles proved that only the gas puffing system would be insufficient and that deuterium pellets are needed for the DTT fuelling. Neutron rates were evaluated and found compatible with the present design of the neutron shields. This PhD modelling work led to the optimisation of the device size and of the reference heating mix, and provided reference profiles for diagnostic system design, estimates of neutron yields, calculations of fast particle losses, gas puffing and/or pellet requirements for fuelling, MHD evaluations, and other tasks.
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Severo, José Helder Facundo. "Estudo da rotação de plasma no tokamak TCABR." Universidade de São Paulo, 2003. http://www.teses.usp.br/teses/disponiveis/43/43134/tde-06092012-125249/.
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Este trabalho, que pode ser dividido em duas partes, teórica e experimental, trata da rotação residual de plasma no TCABR. No que se refere à parte teórica, foi obtida uma expressão geral para a velocidade poloidal e o fluxo de calor, para tokamaks com seção transversal arbitrária, em um plasma que está sujeito a um fluxo subsônico toroidal. Foram estudadas em detalhe as dependências da velocidade poloidal com o número de Mach sigma e o fluxo de calor iônico e foi verificado que a velocidade poloidal troca de sentido para um certo valor sigma=sigma IND.0. Também foi verificado que existe um valor sigma=sigma IND.K, a velocidade poloidal começa a diminuir. Quanto ao fluxo de calor, foi observado que ele é fortemente afetado pela geometria e é proporcional a q POT.2, onde q é o fator de segurança. Para q=1, o fluxo de calor tem um máximo para um fator de elongação k=1, correspondente a uma seção transversal circular, diminui com o aumento de k e apresenta um mínimo em k=2. No que se refere à parte experimental,foram obtidos pela primeira vez, no tokamak TCABR, os perfis radiais das velocidade de rotação poloidal e toroidal para um regime colisional, usando o deslocamento Doppler das linhas espectrais das impurezas de CIII (646,74nm) e CVI (529,02nm), medidas com um espectrômetro TH1000 de distância focal 1000mm e dispersão linear de 8 A/mm. Os resultados experimentais mostram que a velocidade poloidal tem um máximo de (4,5 + OU -1,0).10 POT.5cm/s, cujo sentido de deriva diamagnética dos elétrons. Estes resultados mostram uma boa concordância com a teoria neoclássica para a região da coluna r=5-14 cm, enquanto que para r>14 cm os resultados experimentais estão de desacordo com a teoria. No que diz respeito à velocidade de rotação toroidal, ela é oposta à corrente de plasma e tem um valor máximo de (20 + OU -1).10 POT.5cm/s, o que está em razoável concordância com o modelo proposto por ) Kim e Diamond. Foi observado que a velocidade de rotação toroidal troca de sentido em r>16 cm, indicando haver um forte cisalhamento da rotação na borda da coluna de plasma. A partir dos resultados das velocidades poloidal e toroidal e do gradiente de temperatura iônica, foi calculada a componente radial do campo elétrico que resultou negativo em toda a coluna de plasma. Finalmente, estes resultados estão em boa concordância com os resultados obtidos em tokamaks semelhantes ao TCABR. Os resultados experimentais para a velocidade poloidal podem ser bem descritos pela teoria neoclássica de rotação em tokamaks, exceto nas regiões próximas ao limitador. No entanto, ainda não existe uma teoria geral satisfatória para explicar os resultados da rotação toroidal do plasma em tokamaks. Existem teorias interessantes, porém não são aplicáveis ao tokamak TCABRIn the present work we investigated theorically and experimentally the plasma residual rotation in the tokamak TCABR. Using the neoelassical theory, general expressions for the poloidal velocity and heat flux were obtained for tokamaks with arbitrary plasma cross-sections, and subsonic toroidal flows. The dependency of the poloidal velocity and the heat flow with Mach number a were analyzed. It was found that the poloidal velocity changes sign for a ccrtain valuc alpfa = alpha 0, a critical value ak of a exists corresponding to a maximum value of ion poloidal velocity, and that for alpha > alpha k the poloidal velocity is a decreasing function of alpha.
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Bae, Cheonho. "Extension of neoclassical rotation theory for tokamaks to account for geometric expansion/compression of magnetic flux surfaces." Diss., Georgia Institute of Technology, 2012. http://hdl.handle.net/1853/45839.
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An extended neoclassical rotation theory (poloidal and toroidal) is developed from the fluid moment equations, using the Braginskii decomposition of the viscosity tensor extended to generalized curvilinear geometry and a neoclassical calculation of the parallel viscosity coefficient interpolated over collision regimes. Important poloidal dependences of density and velocity are calculated using the Miller equilibrium flux surface geometry representation, which takes into account elongation, triangularity, flux surface compression/expansion and the Shafranov shift. The resulting set of eight (for a two-ion-species plasma model) coupled nonlinear equations for the flux surface averaged poloidal and toroidal rotation velocities and for the up-down and in-out density asymmetries for both ion species are solved numerically. The numerical solution methodology, a combination of nonlinear Successive Over-Relaxation(SOR) and Simulated Annealing(SA), is also discussed. Comparison of prediction with measured carbon poloidal and toroidal rotation velocities in a co-injected and a counter-injected H-mode discharges in DIII-D [J. Luxon, Nucl. Fusion 42, 614 (2002)] indicates agreement to within <10% except in the very edge in the co-injected discharge.
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Hornung, Grégoire. "Etude de la turbulence plasma par réflectrométrie à balayage ultra-rapide dans le tokamak Tore Supra." Thesis, Aix-Marseille, 2013. http://www.theses.fr/2013AIXM4741/document.
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La turbulence plasma engendre un transport anormal de la chaleur et des particules qui dégrade l’efficacité d’un réacteur de fusion. La mesure de la turbulence plasma dans un tokamak est donc essentielle à la compréhension et au contrôle de ce phénomène. Parmi les instruments de mesure à disposition, le réflectomètre à balayage installé sur le tokamak Tore Supra a accès à la densité du plasma et ses fluctuations depuis le bord jusqu’au centre des décharges, avec une excellente résolution spatiale (mm) et temporelle (µs), de l’ordre des échelles de la turbulence. Cette thèse est dédiée à la caractérisation de la turbulence plasma dans Tore Supra à partir de mesures de réflectométrie à balayage ultrarapide. Des analyses de corrélations ont permis d’évaluer les échelles spatiales et temporelles de la turbulence ainsi que sa vitesse radiale. Dans la première partie, la caractérisation des propriétés de la turbulence à partir des profils de densité reconstruits est discutée, notamment au travers d’une comparaison avec les données des sondes de Langmuir. Ensuite une étude paramétrique est présentée mettant en relief l’effet de la collisionalité sur la turbulence, dont une interprétation est proposée en termes de stabilisation d’une turbulence électronique due aux électrons piégés. Finalement, on illustre comment le chauffage additionnel produit une modification locale de la turbulence dans le plasma proche des parois, se traduisant par une augmentation de la vitesse des structures et une diminution de leur temps de corrélation. L’effet supposé des potentiels rectifiés générés par l’antenne est étudié à l’aide de simulationsThe performance of a fusion reactor is closely related to the turbulence present in the plasma. The latter is responsible for anomalous transport of heat and particles that degrades the confinement. The measure and characterization of turbulence in tokamak plasma is therefore essential to the understanding and control of this phenomenon. Among the available diagnostics, the sweeping reflectometer installed on Tore Supra allows to access the plasma density fluctuations from the edge to the centre of the plasma discharge with a fine spatial (mm) and temporal resolution (µs ) , that is of the order of the characteristic turbulence scales.This thesis consisted in the characterization of plasma turbulence in Tore Supra by ultrafast sweeping reflectometry measurements. Correlation analyses are used to quantify the spatial and temporal scales of turbulence as well as their radial velocity. In the first part, the characterization of turbulence properties from the reconstructed plasma density profiles is discussed, in particular through a comparative study with Langmuir probe data. Then, a parametric study is presented, highlighting the effect of collisionality on turbulence, an interpretation of which is proposed in terms of the stabilization of trapped electron turbulence in the confined plasma. Finally, it is shown how additional heating at ion cyclotron frequency produces a significant though local modification of the turbulence in the plasma near the walls, resulting in a strong increase of the structure velocity and a decrease of the correlation time. The supposed effect of rectified potentials generated by the antenna is investigated via numerical simulations
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Song, Shaodong. "Etude du transport de la chaleur et des particules dans les tokamaks Tore Supra et HL-2A." Thesis, Aix-Marseille 1, 2011. http://www.theses.fr/2011AIX10142/document.
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Le transport de la chaleur et des particules est un des sujets de recherche fondamentaux de la physique des plasmas chauds confinés par des champs magnétiques, systèmes physiques qui sont étudiés dans le cadre des recherches sur la fusion thermonucléaire contrôlée. Ces phénomènes de transport sont essentiellement liés à la turbulence électromagnétique et ils sont donc extrêmement difficiles à modéliser par la théorie. Des expériences spécifiques sont alors réalisées sur des machines expérimentales, telles que les tokamaks ou les stellarators, afin d'améliorer la connaissance de ces phénomènes. Cette thèse décrit des études expérimentales de ce type réalisées sur deux tokamaks de grande dimension: Tore Supra (machine basée au CEA/Cadarache) et HL-2A (basée au South-Western Institute of Physics, Chengdu, Chine). La technique utilisée consiste à injecter, de façon modulée dans le temps, des ondes de forte puissance afin de perturber la température électronique du plasma, et des faisceaux supersoniques de particules pour en perturber la densité. La température est mesurée par l'Emission Cyclotronique Electronique et la densité par Réflectométrie micro-onde. Ces expériences on mis en évidence une convection de la chaleur vers l'intérieur du plasma (un phénomène dont l'existence est toujours controversée), et des effets dus aux termes non-diagonaux de la matrice de transport. Ces résultats ont été comparés aux modèles de transport existantsHeat and particle transport is one of the fundamental subjects of research in the physics of hot plasmas confined by magnetic fields, a class of physical systems that are studied in the framework of research on controlled thermonuclear fusion. These transport phenomena are mainly related to electromagnetic turbulence and are therefore extremely difficult to model at a first-principle level. Specific experiments in this area, on plasma devices such as tokamaks or stellarators, are widely used to improve understanding of these phenomena. This thesis reports on experimental studies performed on two large tokamaks : Tore Supra (based at CEA/Cadarache, France) and HL-2A (based at the South-Western Institute of Physics, Chengdu, China). The technique used consists in modulated injection of wave power to perturb the electron temperature and/or of Supersonic Molecular Beams to perturb the plasma density. Temperature is then measured by Electron Cyclotron Emission and density by Reflectometry, and Fourier analysis is used to determine the transport properties. Evidence has been found of inward heat convection (a phenomenon whose existence is still controversial) as well as of peculiar effects due to the non-diagonal terms of the transport matrix. Comparison with transport models has been carried out
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Mercadier, Laurent. "Spectrocopie de plasma induit par laser pour l'analyse des composants face au plasma de tokamaks : étude paramétrique et mesures autocalibrées." Thesis, Aix-Marseille 2, 2011. http://www.theses.fr/2011AIX22071/document.
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Lors du fonctionnement d'un réacteur de fusion nucléaire par confinement magnétique comme ITER, une fraction de tritium est piégée par les composants face au plasma et doit être mesurée pour des raisons de sureté nucléaire. La spectroscopie de plasma induit par laser est proposée pour effectuer cette mesure. Le plasma laser produit sur des tuiles de Tore Supra en composite à fibre de carbone est analysé à l'aide d'une étude paramétrique : il doit avoir une température supérieure à 10000 K et une densité électronique supérieure à 10^17 cm^-3 pour optimiser l'application. Une méthode "autocalibrée" prenant en compte l'auto-absorption des raies est utilisée pour déterminer la concentration relative d'hydrogène à partir des spectres expérimentaux. La caractérisation spatio-temporelle du panache d'ablation révèle la présence d'un gradient de température dirigé du centre vers la périphérie du plasma. La prise en compte de ce gradient permet de déduire le rapport des concentrations H/C. L'incertitude de la mesure est évaluée et discutée. La mesure du rapport isotopique D/H sous pression réduite d'argon met en évidence un effet de ségrégation qui doit être pris en compte afin d'éviter des erreurs de mesure de l'ordre de 50%. Les matériaux à base de tungstène sont analysés et les difficultés associées aux données spectroscopiques sont abordées. Enfin, la faisabilité de l'analyse LIBS résolue en profondeur est validée pour des échantillons métalliques multicouches préalablement étalonnésDuring the operation of a nuclear fusion device like the future reactor ITER, a fraction of tritium is trapped in the plasma facing components and has to be measured in order to fulfill nuclear safety requirements. Laser-induced breakdown spectroscopy is proposed to achieve this measurement. The laser plasma produced on carbon fibre composite tiles from the Tore Supra reactor is analyzed via a parametric study : it has to have a temperature over 10000 K and an electron density over 10^17 cm^-3 to optimize the application. A calibration-free procedure that takes into account self-absorption is proposed to determine the relative concentration of hydrogen from the experimental spectra. The time- and space-resolved spectral emission of the plasma plume is investigated and reveals the presence of a temperature gradient from the core towards the periphery. This gradient is taken into account and the H/C concentration ratio is deduced. The accuracy of the results is evaluated and discussed. The study of the D/H isotopic ratio under low pressure argon reveals the presence of plume segregation that leads to an error of about 50%, error that can partially be reduced. Tungsten materials are investigated and difficulties related to spectroscopic databases are discussed. Finally, the feasibility of LIBS analysis with depth resolution is validated for multilayered metallic samples
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Breton, Sarah. "Tungsten transport in a tokamak : a first-principle based integrated modeling approach." Thesis, Aix-Marseille, 2018. http://www.theses.fr/2018AIXM0007.
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La fusion par confinement magnétique est actuellement la voie la plus avancée pour produire de l’énergie grâce à la réaction de fusion. L’un des défis à relever concerne la contamination du plasma par le Tungstène (W), un matériau capable de résister aux hauts flux de chaleur. A cause de son grand nombre atomique, le W rayonne dans les plasmas de tokamak. S’il s’accumule au cœur du tokamak, il refroidit le plasma. Il est donc crucial de comprendre les mécanismes du transport du W et d’identifier les paramètres favorisant son accumulation. Le W interagit de façon non-linéaire avec les différents paramètres du plasma. La simulation intégrée est le seul outil permettant à tous ces paramètres d’être simulés de façon auto-consistante durant plusieurs temps de confinement. Pour la première fois, l’outil de simulation intégrée est couplé à des codes de transport premiers principes modélisant de façon auto-consistante les transports turbulent et collisionnel du W, les profils de densité, température, rotation, radiation, et l’évolution du chauffage. Pour des raisons numériques, certains phénomènes ne sont pas modélisés et l'interaction plasma/paroi interne est simplifiée. A chaque pas de temps, cette simulation reproduit avec succès les signaux expérimentaux et le comportement du W. De plus, des acteurs responsables de l’accumulation du W (la rotation et la source centrale de particules) sont identifiés. Enfin, la simulation intégrée a permis de mettre en lumière l’effet stabilisant du W sur la turbulence. Le travail accompli montre que la simulation intégrée premiers principes permet désormais d'optimiser à l'avance les scénarios de plasma afin d'y limiter l'accumulation de WMagnetic confinement fusion is currently the most advanced way to produce energy thanks to Deuterium/Tritium reaction. One of the challenges is the limitation of the reaction contamination because of Tungsten (W), a material capable of resisting high heat fluxes. W large atomic number causes W to radiate inside tokamak plasmas. If W accumulates in the central part, it cools down the plasma. It is therefore crucial to understand the mechanisms of W transport and identify the actuators of the accumulation process. W transport is involved in complex interplays with the plasma parameters (density, temperature, rotation). Therefore the use of integrated modeling is mandatory in order to evolve self-consistently all those parameters for several confinement times. For the first time, an integrated modeling tool is coupled to first-principle transport codes to self-consistently simulate the time evolution of the W behavior, as well as the evolution of density, temperature, rotation profiles, radiation and external heating. For numerical reasons, several phenomena are not modeled, and the physics of the interaction with the inner wall is simplified. At each time step, this simulation successfully reproduces experimental profiles and the W central accumulation. Moreover, actuators of the central W accumulation (rotation and central particle fueling) were identified. Finally, integrated modeling simulation allowed bringing out a very interesting non-linear mechanism: the stabilizing effect of W on turbulence. This work demonstrates that first-principles integrated modeling now allows to design and optimize in advance plasma scenarios with limited W central accumulation
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Lerche, Ernesto Augusto. "Aquecimento do plasma por ondas de Alfvén no tokamak TCABR." Universidade de São Paulo, 2003. http://www.teses.usp.br/teses/disponiveis/43/43134/tde-17072012-141903/.
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Os resultados de uma extensa campanha experimental, realizada no tokamak TCABR, para se investigar a física das ondas de Alfvén e suas aplicações para o aquecimento de plasmas em tokamaks são apresentados. Ao longo das investigações, foram testados dois tipos de antena, tendo sido observado aquecimento considerável do plasma com ambas, mesmo com valor moderado da potência RF injetada no plasma. Diversas configurações de excitação e diversas condições do plasma foram investigadas, e foi verificado que a escolha correta da helicidade da onda excitada PE crucial para se reduzir o acoplamento parasítico com o plasma periférico. Também foi verificada a importância de uma limpeza periódica da superfície das antenas, realizada durante as descargas de limpeza do tokamak, para melhorar o desempenho dos experimentos com aquecimento por ondas de Alfvén. Com a antena original, que produz um espectro poloidal bastante selecionado, a tensão de polarização dinâmica induzida nas antenas observada durante os experimentos era alta, aumentando a taxa de sputtering em seus elementos e podendo, inclusive, levar à disruptura do plasma em potêncis RF mais elevadas. Com o novo tipo de antena, projetado com dimensões poloidais reduzidas, a tensão de polarização induzida caiu pela metade. No entanto, o acoplamento parasítico com a borda do plasma aumentou, como foi indicado por maiores perturbações observadas nos potenciais do SOL, nesse caso. Ademais, a taxa de injeção/ionização de impurezas parece ser maior do que a observada com a antena original em condições semelhantes, como foi indicado pór um aumento maior no sinal do bolômetro durante o pulso RF e por medidas de espectroscopia. Esses fatos sugerem que o espectro excitado pela antena nova é menos seletivo quanto à componente poloidal M, e os modos eletrostáticos devem estar sendo excitados com amplitude considerável. As modificações causadas pela absorção das ondas de Alfvén no perfil radial da temperatura eletrônica do plasma puderam ser estudadas com um radiômetro heteródino de varredura ECE. Esses estudos nos permitiram determinar experimentalmente os perfis radiais de deposição de potência RF no plasma, que estão em surpreendente concordância com os perfis de deposição de potência RF no plasma, que estão surpreendente concordância com os perfis de deposição teóricos, calculados com um código cinético-toroidal para as condições típicas do TCABR. Esses resultados são inéditos em pesquisas com ondas de Alfvén, e reforçam a sua utilização para aquecimento localizado de plasmas e controle de fluxos cizalhados em tokamaks.The results of na extensive experimental campaign performed in the TCABR tokamak to investigate the Physics of the Alfvén wave and its application to tokamak plasma heating are presented. In the course of the experiments, Téo types of Alfvén Wave antennae were studied, and considerable plasma heating was observed in both cases, even with rather small amount of RF Power injected in the plasma. Many antennae configurations and plasma conditions were tried out, and it was verified that the correct choice of the helicity of the excited wave is crucial to reduce the parasitic coupling with the edge plasma. It was also noticed that periodic conditioning of the antenna surface, performed together with the daily tokamak cleaning discharges, also contributes to improve the performance of the heating experiments. With the first antenna type, which produced a rather well defined poloidal spectrum, the dynamic polarication voltage induced in the antennae during the RF experiments was high, causing increased sputtering of its elements and, for higher RF powr input, even plasma disruptions. With the new antenna type, designed with smaller poloidal dimensions, the dynamic polarization voltage of the antenna was reduced twice. However the parasitic coupling with the plasma hás increased, as indicated by stronger perturbations of the electrostatic potentials in the scrape-off layer observed in this case. In addition, the impurity injection/ionization rate also seems to have increased with respect to the previous antenna type in approximately the same conditions, as indicated by a stronger rise in the bolometer signal observed during the RF pulse, and by spectroscopic measurements. These facts suggest that, with the new antenna type, the excited wave spectrum is rather broad with respect to the poloidal wave number M, and electrostatic modes must be excited with quite high amplitude. The change in the radial profiles of the electron temperature due to the Alfvén wave absorption could be studied with a heterodyne sweping ECE radiometer. These sutidies allowed us to determine experimentally the RF Power deposition profiles inside the plasma, which were in surprisingly good agreement with the theoretical deposition profiles, calculated with a kinetic-toroidal code for the TCABR plasma conditions. These results are unprecedented in experimental Alfvén wave research, and strengthen the use of these waves for localized plasma heating and shear flow control in tokamaks.
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Marx, Alain. "Deux étapes majeures pour le développement du code XTOR : parallélisation poussée et géométrie à frontière libre." Thesis, Université Paris-Saclay (ComUE), 2017. http://www.theses.fr/2017SACLX095/document.
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Le code XTOR-2F simule la dynamique 3D des instabilités MHD bi-fluides de plasmas de tokamaks.La première partie de la thèse a été consacrée à la parallélisation du code XTOR-2F. Le code a été parallélisé significativement malgré la représentation pseudo-spectrale pour les deux directions angulaires, la raideur des équations résolues et l’utilisation d’une décomposition LU exacte afin d’inverser le préconditionneur physique. Le temps d’exécution de la version parallèle est un ordre de grandeur plus petit que la version séquentielle sur un maillage basse résolution. L’accélération croît ensuite avec la taille du maillage. La parallélisation permet également de réaliser des simulations avec des maillages plus grands, autrefois non réalisables par la limitation du stockage en RAM.La seconde partie de la thèse a été consacrée au développement d’une version du code permettant de réaliser des simulations en géométrie à frontière libre, s’approchant de la géométrie des tokamaks expérimentaux de grandes tailles. Les conditions initiales sont fournies par le code d’équilibre CHEASE à l’intérieur du plasma. A l’extérieur du plasma, la solution a été étendue en ajustant le potentiel magnétique avec un ensemble de bobines magnétiques poloïdales externes. Les conditions de bord utilisent des fonctions de Green afin de calculer une matrice de transfert permettant de relier les composantes tangentes et normales du champ magnétique externe à la coque avec la solution interne. Ceci permet de modéliser une coque résistive fine. Cette nouvelle version élargie le domaine d’investigation de XTOR-2F, autrefois restreint aux instabilités internes, aux instabilités externes. Le comportement linéaire du code est validé sur deux familles d’instabilités, les modes axisymétriques n = 0 et les kinks externes n = 1 / m = 2. Afin de valider le comportement non linéaire, des simulations en MHD résistive de modes tearing à bêta nul évoluant vers un état stationnaire ont été réaliséesThe XTOR-2F code simulates the 3D dynamics of full bi-fluid MHD instabilities in tokamak plasmas.The first part of the thesis was dedicated to the parallelisation of XTOR-2F code. The code has been parallelised significantly despite the numerical profile of the problem solved, i.e. a discretisation with pseudo-spectral representations in all angular directions, the stiffness of the two-fluid stability problem in tokamaks, and the use of a direct LU decomposition to invert the physical pre-conditioner. The execution time of the parallelised version is an order of magnitude smaller than the sequential one for low-resolution cases, with an increasing speedup when the discretisation mesh is refined. Moreover, it allows to perform simulations with higher resolutions, previously forbidden because of memory limitations.The second part of the thesis was dedicated to the development of free boundary condition. The original fixed boundary computational domain of the code was generalised to a free-boundary one, thus approaching closely the geometry of today’s and future large experimental devices. The initial conditions are given by the CHEASE equilibrium code inside the plasma. Outside the plasma, fitting the magnetic potential at the CHEASE computation domain boundary with a set of external poloidal magnetic coils extends the solution. The boundary conditions use Green functions to construct a response matrix matching the normal and tangential components of the outside magnetic field with the inside solution. A thin resistive wall can be added to the computational domain. This new numerical setup generalises the investigation field from internal MHD instabilities towards external instabilities. The code linear behaviour is validated with two families of instabilities, n = 0 axisymmetric modes and n = 1/m = 2 external kinks. In order to validate the nonlinear behaviour, nonlinear resistive MHD simulations of tearing modes at zero beta evolving to a stationary state have been performed
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Morel, Pierre. "Le modèle "water bag" appliqué aux équation cinétiques des plasmas de Tokamak." Phd thesis, Université Henri Poincaré - Nancy I, 2008. http://tel.archives-ouvertes.fr/tel-00453088.
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Ce travail a porté sur l'étude des instabilités de gradient de température ioniques (ITG) en géométrie cylindrique, le champ magnétique étant supposé constant et dirigé selon l'axe du cylindre. Une fonction de distribution discrète en forme de marche d'escalier est utilisée pour décrire la direction de vitesse parallèle au champ magnétique. L'équation de Vlasov se résume à un système de type multi fluides couplés par l'équation de quasi neutralité. Chaque fluide est décrit par un système fermé d'équations (continuité, Euler et fermeture adiabatique), caractéristiques d'un fluide incompressible, d'ou la dénomination de sac d'eau ou “water bag”. Le recours à cette description water bag est particulièrement intéressant dans le cas de problèmes à une seule dimension en vitesse. Ainsi, dans le cas des plasmas fortement magnétisés, un modèle water bag peut se combiner avantageusement aux modèles dits girocinétiques. Les paramètres associés a la représentation water bag ont pu être identifiés et reliés aux grandeurs macroscopiques par le biais d'une méthode originale d'équivalence au sens des moments. L'analyse water bag des ITG a permis de valider le modèle et les méthodes choisies. Ce travail a également permis de montrer que le concept de water bag peut sans problème prendre en compte des effets variés comme ceux liés a l'introduction d'un rayon de Larmor fini, tout comme à la description d'un plasma composé de plusieurs espèces d'ions.
Books on the topic "Plasmi in tokamak"
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Pitcher, Charles Spencer. Tokamak plasma interaction with limiters. Downsview, Ont: Institute for Aerospace Studies, 1988.
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Pitcher, C. S. Tokamak plasma interaction with limiters. Mississauga, Ont: Ontario Hydro, 1987.
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Kadomt͡sev, B. B. Tokamak plasma: A complex physical system. Bristol, UK: Institute of Physics Pub., 1992.
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Kloe, J. De. Pellet-plasma interaction in a tokamak. Eindhoven: University of Eindhoven, 2000.
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Theory of tokamak plasmas. Amsterdam: North-Holland, 1989.
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Emaami-Khonsaari, Majid. Modelling and control of plasma position in the STOR-M Tokamak. Saskatoon, Sask: Plasma Physics Laboratory, University of Saskatchewan, 1990.
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Zhang, Wei. Improved confinement and edge plasma fluctuations in the STOR-M tokamak. Saskatoon, Sask: University of Saskatchewan, Plasma Physics Laboratory, 1991.
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Zhang, Wei. Plasma auto-biasing during Ohmic H-mode in the STOR-M tokamak. Saskatoon, Sask: University of Saskatchewan, Plasma Physics Laboratory, 1993.
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Mitarai, Osamu. Alternating current plasma operation in the STOR-M tokamak. Saskatoon, Sask: Plasma Physics Laboratory, University of Saskatchewan, 1995.
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Jain, K. K. Measurement of plasma rotation velocities with electrode biasing in the Saskatchewan Torus-Modified (STOR-M) Tokamak. Saskatoon, Sask: Plasma Physics Laboratory, University of Saskatchewan, 1994.
Find full textBook chapters on the topic "Plasmi in tokamak"
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Miyamoto, Kenro. "Tokamak." In Plasma Physics for Controlled Fusion, 337–88. Berlin, Heidelberg: Springer Berlin Heidelberg, 2016. http://dx.doi.org/10.1007/978-3-662-49781-4_15.
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Coccorese, E., and F. Garofalo. "Plasma Position Control." In Tokamak Start-up, 337–51. Boston, MA: Springer US, 1986. http://dx.doi.org/10.1007/978-1-4757-1889-8_22.
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Seidel, U., K. Lackner, G. Lappus, H. Preis, and H. Woyke. "Plasma Position Control in ASDEX Upgrade." In Tokamak Start-up, 325–36. Boston, MA: Springer US, 1986. http://dx.doi.org/10.1007/978-1-4757-1889-8_21.
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Hastie, R. J. "Sawtooth Instability in Tokamak Plasmas." In Plasma Physics, 177–204. Dordrecht: Springer Netherlands, 1998. http://dx.doi.org/10.1007/978-94-011-4758-3_13.
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Ehst, David A., and Kenneth Evans. "Plasma Current Profile Shaping with RF-Current Drive." In Tokamak Start-up, 269–80. Boston, MA: Springer US, 1986. http://dx.doi.org/10.1007/978-1-4757-1889-8_17.
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Ariola, Marco, and Alfredo Pironti. "Plasma Vertical Stabilization." In Magnetic Control of Tokamak Plasmas, 101–15. Cham: Springer International Publishing, 2016. http://dx.doi.org/10.1007/978-3-319-29890-0_7.
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Kikuchi, Mitsuru, and Masafumi Azumi. "Plasma Equilibrium in Tokamak." In Frontiers in Fusion Research II, 17–44. Cham: Springer International Publishing, 2015. http://dx.doi.org/10.1007/978-3-319-18905-5_2.
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Sharapov, Sergei. "Equilibrium of Tokamak Plasma." In Energetic Particles in Tokamak Plasmas, 39–47. First edition. | Boca Raton : CRC Press, 2021.: CRC Press, 2021. http://dx.doi.org/10.1201/9781351002820-4.
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Ariola, Marco, and Alfredo Pironti. "Plasma Magnetic Control Problem." In Magnetic Control of Tokamak Plasmas, 77–90. Cham: Springer International Publishing, 2016. http://dx.doi.org/10.1007/978-3-319-29890-0_5.
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Wagner, F., and K. Lackner. "Divertor Tokamak Experiments." In Physics of Plasma-Wall Interactions in Controlled Fusion, 931–1004. Boston, MA: Springer US, 1986. http://dx.doi.org/10.1007/978-1-4757-0067-1_21.
Full textConference papers on the topic "Plasmi in tokamak"
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Rowan, William L. "Spectroscopic diagnostics and atomic physics experiments in tokamak plasmas." In OSA Annual Meeting. Washington, D.C.: Optica Publishing Group, 1988. http://dx.doi.org/10.1364/oam.1988.thk2.
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As for other tokamaks, spectroscopy of the plasma produced in the Texas Experimental Tokamak (TEXT) has resulted in significant contributions to both atomic and plasma physics. The identification of magnetic and electric dipole transitions in highly ionized species allowed the determination of energy levels while increasing the availability of emission lines for plasma experiments and for use as standards in other atomic physics studies. Collisional rate measurements provided benchmarks for atomic theory and the basic data for implementation of some plasma diagnostics. Complementing the usual plasma physics role as a tool in the study of impurity behavior or ion energy transport, spectroscopy is now also employed in the study of plasma turbulence, MHD stability, and the nonthermal effects of rf heating. As accuracy requirements for some of these measurements increase, active spectroscopic techniques are displacing the familiar passive ones. The experiments which illustrate this review were all conducted on TEXT and are the work of physicists from a number of laboratories. They are chosen to emphasize advances in physics and to describe the critical role of spectroscopy in plasma physics.
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Tarkeshian, R., M. Ghoranneviss, K. Salem, A. Talebi Taher, P. Khorshid, S. M. Atyabi, Hans-Jürgen Hartfuss, Michel Dudeck, Jozef Musielok, and Marek J. Sadowski. "Different Methods for Measuring Plasma Displacement in Tokamaks, Construction & Compensation of Continuous Coils in IR-T1 Tokamak." In PLASMA 2007: International Conference on Research and Applications of Plasmas; 4th German-Polish Conference on Plasma Diagnostics for Fusion and Applications; 6th French-Polish Seminar on Thermal Plasma in Space and Laboratory. AIP, 2008. http://dx.doi.org/10.1063/1.2909116.
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Waldon, C., R. Morrell, D. Buckthorpe, M. Davies, and P. Sherlock. "Engineering Practices for Tokamak Window Assemblies." In 17th International Conference on Nuclear Engineering. ASMEDC, 2009. http://dx.doi.org/10.1115/icone17-75858.
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For fusion tokamak reactors the diagnostics and RF heating systems require the use of components with parts made of non-metallic materials. These can form part of the vacuum boundary of the tokamak which is the primary safety boundary and have a function of containing tritium fuel or activated gases and particulate debris. The engineering practices for such components and non-metallic materials are in an early state of preparation and require development to enable systems to be used in a safety and licensing context. Such developments will have to reflect the brittle nature of the materials, and are likely to be based on established arguments developed within the nuclear industry, such as containment and defence in depth. Given these requirements this task is a major challenge. The window systems fall broadly into two categories: • Transmission windows for the input of high-power microwaves to drive and heat the plasma; • Diagnostic windows to monitor the plasma. Currently there are no established fusion design codes that can be used to assure nuclear safety and a consistent engineering approach for either application. This paper reviews the progress made in developing such practices for transmission and diagnostic windows made from ceramic materials. The investigations undertaken and the engineering practices addressed for the tokamak windows generally fall into the following areas: • reviews of potential candidate materials along with a summary of the available property data; • definition of the function of torus window assemblies and an outline of the complexity and variety of design considerations (including historical failures, and statutory requirements); • development of the design methodology for technical ceramics; • definition of the design routes considered and selected (rule, analysis, experiment); • consideration of the material data available (or lack of) for technical ceramics and their failure criteria; • qualification and design of metallic / ceramic joints; • definition of the requirements with regard to quality control, from manufacture to in-service inspection; • development and formation of a draft code procedure. The practices and procedures developed are considered to be an important contribution and significant step forward in the development of a fusion tokamak windows code. Important contributions have been made to the design, procurement and installation philosophies for windows, especially the development of design criteria and the application of pressure proof-testing. This paper provides a review of key requirements and issues, with recommendations to allow development of the code for acceptance by nuclear regulators for tokamaks such as the International Tokamak Experimental Reactor (ITER) and future fusion reactor power plants.
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Suckewer, S., L. Bromberg, and D. Cohn. "Small Scale Tokamak for X-Ray Lithography." In Soft X-Ray Projection Lithography. Washington, D.C.: Optica Publishing Group, 1992. http://dx.doi.org/10.1364/sxray.1992.wb3.
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A toroidal plasma device (tokamak) with electron temperature in the range of 150-200 eV and density ~1013 particles/cm3 can be built as a very compact and relatively inexpensive machine (~3 M$). A tokamak with a major radius R ≈ 1m, minor radius r ≈ 0.1m, and confining magnetic field ~ 5k Gauss is not a very attractive for fusion research, however it can be an excellent source of soft X-ray radiation. In particular if operated in a steady state regime or at a high repetition rate it can provide several orders of magnitude more soft X-ray radiation than a small synchrotron with an undulator. This can be seen easily by comparing the total radiated power of a small tokamak and a small sychrotron, taking into account the spectral intensity distribution of line radiation from the tokamak plasma and of continuum radiation from synchrotron. We will present related calculations. Based on the calculations we will discuss the usefulness of a small tokamak for X-ray projection and proximity lithography and simple methods to change the dominant lines in the plasma radiation spectrum.
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Walker, Michael L., Peter De Vries, Federico Felici, and Eugenio Schuster. "Introduction to Tokamak Plasma Control." In 2020 American Control Conference (ACC). IEEE, 2020. http://dx.doi.org/10.23919/acc45564.2020.9147561.
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Kim, C. S., H. S. Lee, and M. Kwon. "KSTAR Tokamak Neutronic Analysis." In IEEE Conference Record - Abstracts. 2005 IEEE International Conference on Plasma Science. IEEE, 2005. http://dx.doi.org/10.1109/plasma.2005.359096.
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Diebold, D., N. Hershkowitz, and J. Sorensen. "Phaedrus-T tokamak probe measurements." In International Conference on Plasma Sciences (ICOPS). IEEE, 1993. http://dx.doi.org/10.1109/plasma.1993.593411.
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Roberto, M. "Reconnection Bifurcation in Tokamaks." In PLASMA PHYSICS: 11th International Congress on Plasma Physics: ICPP2002. AIP, 2003. http://dx.doi.org/10.1063/1.1594033.
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Cui, Zhiqiang. "Energy Calibration of Scintillator Detectors in Different Neutron Diagnostic System on Tokamak." In 2018 26th International Conference on Nuclear Engineering. American Society of Mechanical Engineers, 2018. http://dx.doi.org/10.1115/icone26-81190.
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The purpose of tokamak plasma diagnostics is to provide the necessary parameters for device protection, operation, and maintenance. It can also supply parameters for fusion physics research. As one of the main ways to diagnose nuclear fusion plasma, neutron diagnosis focuses on the detection of neutrons, produced by the D-D and D-T fusion reactions, to obtain the physical information of internal plasma. Neutron measurements are widely performed on tokamak to provide the essential information on the neutron yield rate of the plasma that is related to fusion power. Since neutron has no electric charge, neutron can’t be ionized directly by the interaction of electrons in the detection material. The interactions between neutron and nuclei, such as nuclear reaction and nuclear recoil, are used to detect neutrons. According to the front sensitive materials, neutron detectors can be divided into gas detectors, scintillator detectors, semiconductor detectors, ionization chambers and so on. Since the magnetic field surrounding Tokamak can have a magnificent influence on the performance of photo-electronic multiplier tubes (PMTs), it is necessary to employ magnetic shielding in designing detectors, thus guaranteeing the proper operation of detectors within a strong magnetic field. Although the PMTs are equipped with magnetic shielding materials by manufacturers, they can only resist the influence of geomagnetic field. Besides the magnetic shielding and neutron/gamma shielding, neutron detectors should be calibrated before used on the tokamak. Nine similar detectors were assembled and calibrated in this paper. The basic idea of processing calibration data is that we should adjust the resolution and the light response function in order to make experiment spectrum and simulation spectrum fit on the recoil proton edge. A special explication is given to the data processing of neutron calibration, followed by an analysis of its resulting light response function and by comparison with PTB’s results.
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Stambauch, R. D., G. Bateman, M. G. Bell, D. Cohn, P. Colestock, R. Goldston, S. C. Jardin, et al. "The compact ignition tokamak (CIT)." In 1990 Plasma Science IEEE Conference Record - Abstracts. IEEE, 1990. http://dx.doi.org/10.1109/plasma.1990.110576.
Full textReports on the topic "Plasmi in tokamak"
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R. Kaita, R. Majeski, R. Doerner, G. Antar, M. Baldwin, R. Conn, P. Efthimion, et al. Liquid Lithium Limiter Effects on Tokamak Plasmas and Plasma-Liquid Surface Interactions. Office of Scientific and Technical Information (OSTI), October 2002. http://dx.doi.org/10.2172/809848.
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Hernandez, J. V., Z. Lin, W. Horton, A. Vannucci, and S. C. McCool. Neural net prediction of tokamak plasma disruptions. Office of Scientific and Technical Information (OSTI), October 1994. http://dx.doi.org/10.2172/10193739.
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Author, Not Given. (High beta tokamak research and plasma theory). Office of Scientific and Technical Information (OSTI), January 1990. http://dx.doi.org/10.2172/6156973.
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Medley, S. S., and K. M. Young. Plasma diagnostics for the compact ignition tokamak. Office of Scientific and Technical Information (OSTI), June 1988. http://dx.doi.org/10.2172/7231452.
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Furth, H. P. Three novel tokamak plasma regimes in TFTR. Office of Scientific and Technical Information (OSTI), October 1985. http://dx.doi.org/10.2172/5109719.
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Singer, C. E., L. P. Ku, and G. Bateman. Plasma transport in a compact ignition tokamak. Office of Scientific and Technical Information (OSTI), February 1987. http://dx.doi.org/10.2172/6685807.
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L.E. Zakharov. Stabilization of tokamak plasma by lithium streams. Office of Scientific and Technical Information (OSTI), August 2000. http://dx.doi.org/10.2172/759425.
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Ward, D. J., and S. C. Jardin. The effects of plasma deformability on the feedback stabilization of axisymmetric modes in tokamak plasmas. Office of Scientific and Technical Information (OSTI), January 1992. http://dx.doi.org/10.2172/5956297.
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Ward, D. J., and S. C. Jardin. The effects of plasma deformability on the feedback stabilization of axisymmetric modes in tokamak plasmas. Office of Scientific and Technical Information (OSTI), January 1992. http://dx.doi.org/10.2172/10115427.
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Hassanein, A., and I. Konkashbaev. Modeling plasma/material interactions during a tokamak disruption. Office of Scientific and Technical Information (OSTI), October 1994. http://dx.doi.org/10.2172/10197014.
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