Orientador: Munemasa Machida / Tese (doutorado) - Universidade Estadual de Campinas, Instituto de Fisica Gleb Wataghin / Made available in DSpace on 2018-08-03T17:00:15Z (GMT). No. of bitstreams: 1 Monteiro_MarcelodeJesusRangel_D.pdf: 3635325 bytes, checksum: 206622522804a8aa49d97c248dffa665 (MD5) Previous issue date: 2003 / Resumo: Foram executados no tokamak NOVA-UNICAMP os trabalhos da tese que consistiram em duas partes distintas: a primeira ligada à implementação de programas e operação da máquina, e a segunda compreendeu a realização do diagnóstico de espalhamento Thomson utilizado para a medida direta da temperatura eletrônica, e indireta, via espalhamento Rayleigh, da densidade eletrônica do plasma. Com relação à primeira parte, foram desenvolvidos os programas para a aquisição e tratamento dos principais dados provenientes do tokamak e estudado (e determinado) durante os trabalhos as melhores condições de operação da máquina. Já com relação à segunda parte, foi implantado o sistema de multipassagem para o feixe do laser de rubi, que permitiu o aumento da potência incidente em torno de seis vezes, e foi utilizado um novo detector multicanal, o XP 1752 da Philips, que é uma matriz de 8 x 8, totalizando 64 fotomultiplicadoras independentes, e que ainda não havia sido testado na realização do espalhamento Thomson multipassagem e multiespacial. Devido ao número limitado de canais para a aquisição dos dados, via conversores analógico / digital (ADC) , foi possível trabalhar com um total de 20 dos 64 canais do detector multicanal, permitindo se fazer pela primeira vez desde que o tokamak chegou ao laboratório, medidas simultâneas de temperatura e densidade eletrônicas em quatro posições espaciais distintas ao longo do raio do plasma e do tempo de descarga do tokamak / Abstract: In this thesis we present two distinct works implemented on the NOVA-UNICAMP tokamak. The first one involved the determination of the best machine operation conditions with the development of software¿s for machine control, data acquisition and analysis which were not available up to the moment. The second part is the development, set up and use of the Thomson scattering diagnostic with multi-pass, multi-channel, and multi-space capability. This diagnostic permitted us to carry out direct measurements of electron temperature and indirect measurements of electron density using Rayleigh scattering, along a full tokamak discharge time in four different radial position inside the plasma. For the multipass set up, the initial ruby laser power has been increased by a factor of six, which permitted good precision electron temperature and density measurements. The multichannel detection capability has been possible due to the newly available 8 x 8 matrix multi-channel detector XP 1752 from Philips, which, up to our knowledge was used for the first time. Due to the limited numbers of ADC modules available at the time of the measurements, in our Lab., we could use only 20 of the 64 channels which permitted us to perform four spatial temperature measurements during one tokamak discharge. Nevertheless the capability of full spatial electron temperature and density measurements along the tokamak discharges, with good precision, has been demonstrated. / Doutorado / Física / Doutor em Ciências

Orientador: Munemasa Machida / Dissertação (mestrado) - Universidade Estadual de Campinas, Instituto de Fisica Gleb Wataghin / Made available in DSpace on 2018-09-11T21:04:46Z (GMT). No. of bitstreams: 1 Kaminishikawahara_CelsoOssamu_M.pdf: 1343678 bytes, checksum: 01d8d9e063e8f65d41507475b4fca60e (MD5) Previous issue date: 2004 / Resumo: O objetivo deste trabalho foi de implementar um sistema de diagnóstico na borda do plasma, na Região da Sombra do Limitador (RSL), do tokamak Nova-Unicamp, até então inexistente. O sistema de diagnóstico desenvolvido foi a sonda de Langmuir com a qual pudemos observar as variações temporais e espaciais da temperatura e densidade eletrônica na borda do plasma. Para a implementação desse sistema de diagnóstico no nosso tokamak foram construídas duas sondas de Langmuir , e desenvolvidas fontes de tensões de ondas senoidal e dente-de-serra, para alimentar essas sondas. Para o tratamento dos dados obtidos com a fonte de tensão dente-de-serra, foi desenvolvido um programa na linguagem C chamado Programa Langmuir, para uma determinação rápida da temperatura e densidade a cada disparo em descarga do tokamak. Com esse sistema implantado determinamos que a temperatura e a densidade eletrônica na interface entre o limitador e o plasma confinado, são 8,7 eV e 5,6x10 12cm -3respectivamente. E que o crescimento da densidade é mais abrupto que o da temperatura.Observamos também que as variações temporal da temperatura e densidade eletrônica ao longo da descarga mantiveram-se constantes nos valores de 8eV e 4x10 12cm -3respectivamente, na posição a 2mm atrás do limitador / Abstract: The objective of this work was to implement a system of diagnostic near the edge of the plasma, in the scrape-off layer (SOL) of tokamak Nova-Unicamp, up to time inexistent. The diagnostic system developed was the Langmuir probe with which we can observe the time and radial profiles of the electron temperature and density in the edge of the plasma. For the implementation of this diagnostic in our tokamak there have been constructed two Langmuir probes. To feed these probes, we have developed a sweep voltage power supply with sawtooth and sinusoidal waveform generator. For the treatment of the data obtained with sawtooth voltage source generator, it has been developed a program, in C language, named Langmuir Program, for a fast determination of the electron temperature and density to each tokamak discharge. With this system, we have determined that the eletron temperature and density at the interface between the limiter and the confined plasma, are 8.7 eV and 5.6x10 12cm -3 respectively, and that the growth of the density is more abrupt than the temperature. We also have observed that the time variations of the temperature and density, during the plateau of tokamak discharge, had respectively remained constant at the values of 8 eV and 4x10 12cm -3 , in the region 2mm behind the limiter / Mestrado / Física / Mestre em Física

Thesis (Ph.D.)--Massachusetts Institute of Technology, Dept. of Nuclear Engineering, 1981. / MICROFICHE COPY AVAILABLE IN ARCHIVES AND SCIENCE. / Includes bibliographical references. / by Kenneth Edmund Kreischer. / Ph.D.

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 TCABR / In 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.

The primary achievement of this study is the development of a new approach for optimizing the plasma shape in a tokamak fusion energy reactor. In the interest of producing the largest possible fusion power output, the shape is optimized to allow for the highest possible beta - the ratio of the fluid to magnetic pressure - that can be sustained without the onset of magnetohydrodynamic (MHD) instabilities. To this end, the study explores the beta-domain that is stabilizable by bulk plasma rotation, with rotation timescales comparable to the resistive dissipation time of the plasma tearing surfaces or of the surrounding vacuum chamber. Modern feedback control systems are able to apply external magnetic fields which are phased to emulate the effect of plasma rotation, making the technique applicable even to large tokamaks with inadequate plasma rotation. In order to explore how the rotationally stabilizable beta-domain is affected by plasma shaping, a new semi-analytic MHD model of a tokamak has been developed. In addition to shaped toroidal tokamak geometry, the model contains dissipative effects resulting from resistivity in both the plasma and in the vacuum-chamber wall. The inclusion of plasma and wall resistivity introduces a lower beta-limit, associated with the onset of an unstable MHD mode, which can become dominated by either resistive-plasma (tearing) or resistive-wall effects in different parts of the parameter space. The computation time for analyzing the mode stability is greatly reduced by approximating the plasma current to reside in a thin layer, a form known as a sharp-boundary model. With fast calculations that focus on the key physics of these MHD instabilities, the model is able to explore qualitative trends of rotational stabilizability over a broad range of plasma shapes. Results of this study predict that varying the elongation or triangularity of the plasma cross-section can lead to qualitatively different beta-limits for the rotationally stabilizable domain. As the shape is varied, the upper bound in beta for rotational stabilization is found to switch from resistive-wall dominated behavior to resistive-plasma dominated behavior. The optimal plasma shape, associated with the highest beta-limit achievable with plasma rotation, is shown to be at the crossing point between the two domains. This discovery provides a basis for understanding existing experimental results and lays the groundwork for more quantitative studies with larger codes.

This thesis presents simulation and experimental work on feedback control of the \emph{phase} of non-rotating magnetic islands (locked modes) in the DIII-D tokamak, as well as its application to synchronized modulated current drive, for stability studies and control of the locked mode \emph{amplitude}. A numerical model has been developed to predict mode dynamics under the effect of various electromagnetic torques, due to the interaction with induced currents in the wall, error fields, and applied resonant magnetic perturbations (RMPs). This model was adapted to predict entrainment capabilities on ITER, suggesting that small (5~cm) islands can be entrained in the sub-10~Hz frequency range. Simulations and subsequent experiments on DIII-D demonstrated a novel technique to prevent locked modes. Preemptive entrainment applies a rotating RMP before a neoclassical tearing mode fully decelerates such that it will be entrained by the RMP and mode rotation can be sustained. A feedback control algorithm was designed and implemented on DIII-D to offer the ability to prescribe any toroidal phase to the mode and to allow for smoother entrainment. Experimental results confirmed simulation predictions of successful entrainment, and demonstrated one possible application to electron cyclotron current drive (ECCD). Feedback-controlled mode rotation and pre-programmed ECCD modulation were synchronized at DIII-D. This allowed a fine control of the ECCD deposition relative to the island O-point. Experiments exhibited a modulation of the saturated island width, in agreement with time-dependent modeling of the modified Rutherford equation. This work contributes to control and suppression of locked modes in future devices, including ITER.

Electron heat transport at the scale of electron gyroradius are investigated via numerical simulation of a fluid model and a role of E x B shear flow with intermediate E x B shearing rate is explored in Euler's equation. The anomalous transport, enhanced transport due to turbulent electro-magnetic fields caused by plasma instabilities, has been a focus of the inter-national fusion research communities. Among the instabilities, the drift-type instabilities from the pressure-gradient universal in magnetic fusion devicesare considered responsible for the anomalous transport. In the current status of wide use of wave heating on electrons and subsequent high core electron temperature, the turbulent heat loss through electrons has one of the most important science element in preventing the large fusion tokamaks from reaching breakeven in the past decade. The Electron Temperature Gradient fluid model consists of electrostatic potential, toroidal magnetic flux function and electron temperature (or pressure) describing electron drift waves. The fluid model proves to be highly useful to electron heat flux analysis in fusion machines. We analyze the discharges in National Spherical Tokamak eXperiment(NSTX) and Tokamak Configuration Variable (TCV) and found that the electron thermal diffusivities can be explained in terms of the mixing length argument based on electron gyroradius, linear theory and our nonlinear fluid simulation. The nonlinear fluid model predicts reasonable heat flux observed in the experiments. Based on the analysis, we investigate the dependences of the dynamics on the ratio of electron and ion temperature T[subscript e]/T[subscript i] and plasma beta [beta subscript e-]. The nonlinear dynamics such as saturation mechanism of the ETG turbulence and the electromagnetic dynamics in terms of micro-tearing at the scale of electron gyroradius are discussed briegly. In most of plasma confinement devices, the equilibrium radial electric field exists and the turbulence-generated electric field is observed. The coherent structure, called as zonal flow, has been know to be effective to suppress the micro-turbulence. But at intermediate E x B shear, where the vortex eddy turn-over time is comparable to E x B shearing rate, the suppression is weak and the flow shear can leads to vortex amplification through interaction of nonlinear dynamics and shear flow. / text

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