Nonlocal electron heat-flux

Electron thermal conduction in a not quite collisional unmagnetlzed plasma is analysed. The failure of classical results for temperature scale-length up to 100 times larger than thermal mean-free-path for electron scattering, and large ion-charge number Z , is discussed. Recent results from a nonlocal model of conduction at large Z are reviewed. Closed form expressions for Braginskii's coefficients a ,/3 , y for Z =0(1) are derived. An extension of the nonlocal model for Z =0(1) is discussed.

Electron cooling in probe collection from magnetized plasmas with anomalous transport

The electron-retarding range of the current-voltage characteristic of a flat Langmuir probe perpendicular to a strong magnetic field in a fully ionized plasma is analysed allowing for anomalous (Bohm) cross-field transport and temperature changes in the collection process. With probe size and ion thermal gyroradius comparable, and smaller than the electron mean free path, there is an outer quasineutral region with ion viscosity determinant in allowing nonambipolar parallel and cross flow. A potential overshoot lying either at the base or inside the quasineutral region both makes ions follow Boltzmann's law at negative bias and extends the electron-retarding range to probe bias e(j)p ~ +2Too. Electron heating and cooling occur roughly at positive and negative bias, with a re-minimum around efa ~ - 2 7 ^ ; far from the probe heat conduction cools and heats electrons at and radially away from the probe axis, respectively. The potential overshoot with no thermal effects would reduce the electron current Ie, making the In Ie versus 4>p graph downwards-concave,but cooling further reduces Ie substantially, and may tilt the slope upwards past the temperature minimum. The domain of strict validity of our analysis is narrow in case of low ion mass (deuterium), breaking down with the ion Boltzmann law.

Self-similar deflagration in laser half-plasmas

The self-similar motion of a half-space plasma, generated by a linear pulse of laser radiation absorbed anomalously at the critical density, has been studied. The resulting plasma structure has been completely determined for [pulse duration (critical density)maximum irradiation] large enough

Two-temperature model of the coronal irradiated pellets

A two electron-temperature, quasi-steady model of the corona of a laser-ablated pellet is considered. Ablation pressure, critical radius and mass flow rate are determined. Results are close to those obtained with heat flux saturation well below the free-streaming limit.

Dise��o de un m��dulo de IFMIF para irradiaci��n de envolturas regeneradoras basadas en metales l��quidos para reactores de fusi��n

La fusi��n nuclear es, hoy en d��a, una alternativa energ��tica a la que la comunidad internacional dedica mucho esfuerzo. El objetivo es el de generar entre diez y cincuenta veces m��s energ��a que la que consume mediante reacciones de fusi��n que se producir��n en una mezcla de deuterio (D) y tritio (T) en forma de plasma a doscientos millones de grados cent��grados. En los futuros reactores nucleares de fusi��n ser�� necesario producir el tritio utilizado como combustible en el propio reactor termonuclear. Este hecho supone dar un paso m��s que las actuales m��quinas experimentales dedicadas fundamentalmente al estudio de la f��sica del plasma. As�� pues, el tritio, en un reactor de fusi��n, se produce en sus envolturas regeneradoras cuya misi��n fundamental es la de blindaje neutr��nico, producir y recuperar tritio (fuel para la reacci��n DT del plasma) y por ��ltimo convertir la energ��a de los neutrones en calor. Existen diferentes conceptos de envolturas que pueden ser s��lidas o l��quidas. Las primeras se basan en cer��micas de litio (Li2O, Li4SiO4, Li2TiO3, Li2ZrO3) y multiplicadores neutr��nicos de Be, necesarios para conseguir la cantidad adecuada de tritio. Los segundos se basan en el uso de metales l��quidos o sales fundidas (Li, LiPb, FLIBE, FLINABE) con multiplicadores neutr��nicos de Be o el propio Pb en el caso de LiPb. Los materiales estructurales pasan por aceros ferr��tico-martens��ticos de baja activaci��n, aleaciones de vanadio o incluso SiCf/SiC. Cada uno de los diferentes conceptos de envoltura tendr�� una problem��tica asociada que se estudiar�� en el reactor experimental ITER (del ingl��s, ���International Thermonuclear Experimental Reactor���). Sin embargo, ITER no puede responder las cuestiones asociadas al da��o de materiales y el efecto de la radiaci��n neutr��nica en las diferentes funciones de las envolturas regeneradoras. Como referencia, la primera pared de un reactor de fusi��n de 4000MW recibir��a 30 dpa/a��o (valores para Fe-56) mientras que en ITER se conseguir��an <10 dpa en toda su vida ��til. Esta tesis se encuadra en el acuerdo bilateral entre Europa y Jap��n denominado ���Broader Approach Agreement ���(BA) (2007-2017) en el cual Espa��a juega un papel destacable. Estos proyectos, complementarios con ITER, son el acelerador para pruebas de materiales IFMIF (del ingl��s, ���International Fusion Materials Irradiation Facility���) y el dispositivo de fusi��n JT-60SA. As��, los efectos de la irradiaci��n de materiales en materiales candidatos para reactores de fusi��n se estudiar��n en IFMIF. El objetivo de esta tesis es el dise��o de un m��dulo de IFMIF para irradiaci��n de envolturas regeneradoras basadas en metales l��quidos para reactores de fusi��n. El m��dulo se llamar�� LBVM (del ingl��s, ���Liquid Breeder Validation Module���). La propuesta surge de la necesidad de irradiar materiales funcionales para envolturas regeneradoras l��quidas para reactores de fusi��n debido a que el dise��o conceptual de IFMIF no contaba con esta utilidad. Con objeto de analizar la viabilidad de la presente propuesta, se han realizado c��lculos neutr��nicos para evaluar la idoneidad de llevar a cabo experimentos relacionados con envolturas l��quidas en IFMIF. As��, se han considerado diferentes candidatos a materiales funcionales de envolturas regeneradoras: Fe (base de los materiales estructurales), SiC (material candidato para los FCI��s (del ingl��s, ���Flow Channel Inserts���) en una envoltura regeneradora l��quida, SiO2 (candidato para recubrimientos antipermeaci��n), CaO (candidato para recubrimientos aislantes), Al2O3 (candidato para recubrimientos antipermeaci��n y aislantes) y AlN (material candidato para recubrimientos aislantes). En cada uno de estos materiales se han calculado los par��metros de irradiaci��n m��s significativos (dpa, H/dpa y He/dpa) en diferentes posiciones de IFMIF. Estos valores se han comparado con los esperados en la primera pared y en la zona regeneradora de tritio de un reactor de fusi��n. Para ello se ha elegido un reactor tipo HCLL (del ingl��s, ���Helium Cooled Lithium Lead���) por tratarse de uno de los m��s prometedores. Adem��s, los valores tambi��n se han comparado con los que se obtendr��an en un reactor r��pido de fisi��n puesto que la mayor��a de las irradiaciones actuales se hacen en reactores de este tipo. Como conclusi��n al an��lisis de viabilidad, se puede decir que los materiales funcionales para mantos regeneradores l��quidos podr��an probarse en la zona de medio flujo de IFMIF donde se obtendr��an ratios de H/dpa y He/dpa muy parecidos a los esperados en las zonas m��s irradiadas de un reactor de fusi��n. Adem��s, con el objetivo de ajustar todav��a m��s los valores, se propone el uso de un moderador de W (a considerar en algunas campa��as de irradiaci��n solamente debido a que su uso hace que los valores de dpa totales disminuyan). Los valores obtenidos para un reactor de fisi��n refuerzan la idea de la necesidad del LBVM, ya que los valores obtenidos de H/dpa y He/dpa son muy inferiores a los esperados en fusi��n y, por lo tanto, no representativos. Una vez demostrada la idoneidad de IFMIF para irradiar envolturas regeneradoras l��quidas, y del estudio de la problem��tica asociada a las envolturas l��quidas, tambi��n incluida en esta tesis, se proponen tres tipos de experimentos diferentes como base de dise��o del LBVM. ��stos se orientan en las necesidades de un reactor tipo HCLL aunque a lo largo de la tesis se discute la aplicabilidad para otros reactores e incluso se proponen experimentos adicionales. As��, la capacidad experimental del m��dulo estar��a centrada en el estudio del comportamiento de litio plomo, permeaci��n de tritio, corrosi��n y compatibilidad de materiales. Para cada uno de los experimentos se propone un esquema experimental, se definen las condiciones necesarias en el m��dulo y la instrumentaci��n requerida para controlar y diagnosticar las c��psulas experimentales. Para llevar a cabo los experimentos propuestos se propone el LBVM, ubicado en la zona de medio flujo de IFMIF, en su celda caliente, y con capacidad para 16 c��psulas experimentales. Cada c��psula (24-22 mm de di��metro y 80 mm de altura) contendr�� la aleaci��n eut��ctica LiPb (hasta 50 mm de la altura de la c��psula) en contacto con diferentes muestras de materiales. ��sta ir�� soportada en el interior de tubos de acero por los que circular�� un gas de purga (He), necesario para arrastrar el tritio generado en el eut��ctico y permeado a trav��s de las paredes de las c��psulas (continuamente, durante irradiaci��n). Estos tubos, a su vez, se instalar��n en una carcasa tambi��n de acero que proporcionar�� soporte y refrigeraci��n tanto a los tubos como a sus c��psulas experimentales interiores. El m��dulo, en su conjunto, permitir�� la extracci��n de las se��ales experimentales y el gas de purga. As��, a trav��s de la estaci��n de medida de tritio y el sistema de control, se obtendr��n los datos experimentales para su an��lisis y extracci��n de conclusiones experimentales. Adem��s del an��lisis de datos experimentales, algunas de estas se��ales tendr��n una funci��n de seguridad y por tanto jugar��n un papel primordial en la operaci��n del m��dulo. Para el correcto funcionamiento de las c��psulas y poder controlar su temperatura, cada c��psula se equipar�� con un calentador el��ctrico y por tanto el m��dulo requerir�� tambi��n ser conectado a la alimentaci��n el��ctrica. El dise��o del m��dulo y su l��gica de operaci��n se describe en detalle en esta tesis. La justificaci��n t��cnica de cada una de las partes que componen el m��dulo se ha realizado con soporte de c��lculos de transporte de tritio, termohidr��ulicos y mec��nicos. Una de las principales conclusiones de los c��lculos de transporte de tritio es que es perfectamente viable medir el tritio permeado en las c��psulas mediante c��maras de ionizaci��n y contadores proporcionales comerciales, con sensibilidades en el orden de 10-9 Bq/m3. Los resultados son aplicables a todos los experimentos, incluso si son c��psulas a bajas temperaturas o si llevan recubrimientos antipermeaci��n. Desde un punto de vista de seguridad, el conocimiento de la cantidad de tritio que est�� siendo transportada con el gas de purga puede ser usado para detectar de ciertos problemas que puedan estar sucediendo en el m��dulo como por ejemplo, la rotura de una c��psula. Adem��s, es necesario conocer el balance de tritio de la instalaci��n. Las p��rdidas esperadas el refrigerante y la celda caliente de IFMIF se pueden considerar despreciables para condiciones normales de funcionamiento. Los c��lculos termohidr��ulicos se han realizado con el objetivo de optimizar el dise��o de las c��psulas experimentales y el LBVM de manera que se pueda cumplir el principal requisito del m��dulo que es llevar a cabo los experimentos a temperaturas comprendidas entre 300-550��C. Para ello, se ha dimensionado la refrigeraci��n necesaria del m��dulo y evaluado la geometr��a de las c��psulas, tubos experimentales y la zona experimental del contenedor. Como consecuencia de los an��lisis realizados, se han elegido c��psulas y tubos cil��ndricos instalados en compartimentos cil��ndricos debido a su buen comportamiento mec��nico (las tensiones debidas a la presi��n de los fluidos se ven reducidas significativamente con una geometr��a cil��ndrica en lugar de prism��tica) y t��rmico (uniformidad de temperatura en las paredes de los tubos y c��psulas). Se han obtenido campos de presi��n, temperatura y velocidad en diferentes zonas cr��ticas del m��dulo concluyendo que la presente propuesta es factible. Cabe destacar que el uso de c��digos fluidodin��micos (e.g. ANSYS-CFX, utilizado en esta tesis) para el dise��o de c��psulas experimentales de IFMIF no es directo. La raz��n de ello es que los modelos de turbulencia tienden a subestimar la temperatura de pared en mini canales de helio sometidos a altos flujos de calor debido al cambio de las propiedades del fluido cerca de la pared. Los diferentes modelos de turbulencia presentes en dicho c��digo han tenido que ser estudiados con detalle y validados con resultados experimentales. El modelo SST (del ingl��s, ���Shear Stress Transport Model���) para turbulencia en transici��n ha sido identificado como adecuado para simular el comportamiento del helio de refrigeraci��n y la temperatura en las paredes de las c��psulas experimentales. Con la geometr��a propuesta y los valores principales de refrigeraci��n y purga definidos, se ha analizado el comportamiento mec��nico de cada uno de los tubos experimentales que contendr�� el m��dulo. Los resultados de tensiones obtenidos, han sido comparados con los valores m��ximos recomendados en c��digos de dise��o estructural como el SDC-IC (del ingl��s, ���Structural Design Criteria for ITER Components���) para as�� evaluar el grado de protecci��n contra el colapso pl��stico. La conclusi��n del estudio muestra que la propuesta es mec��nicamente robusta. El LBVM implica el uso de metales l��quidos y la generaci��n de tritio adem��s del riesgo asociado a la activaci��n neutr��nica. Por ello, se han estudiado los riesgos asociados al uso de metales l��quidos y el tritio. Adem��s, se ha incluido una evaluaci��n preliminar de los riesgos radiol��gicos asociados a la activaci��n de materiales y el calor residual en el m��dulo despu��s de la irradiaci��n as�� como un escenario de p��rdida de refrigerante. Los riesgos asociados al m��dulo de naturaleza convencional est��n asociados al manejo de metales l��quidos cuyas reacciones con aire o agua se asocian con emisi��n de aerosoles y probabilidad de fuego. De entre los riesgos nucleares destacan la generaci��n de gases radiactivos como el tritio u otros radiois��topos vol��tiles como el Po-210. No se espera que el m��dulo suponga un impacto medioambiental asociado a posibles escapes. Sin embargo, es necesario un manejo adecuado tanto de las c��psulas experimentales como del m��dulo contenedor as�� como de las l��neas de purga durante operaci��n. Despu��s de un d��a de despu��s de la parada, tras un a��o de irradiaci��n, tendremos una dosis de contacto de 7000 Sv/h en la zona experimental del contenedor, 2300 Sv/h en la c��psula y 25 Sv/h en el LiPb. El uso por lo tanto de manipulaci��n remota est�� previsto para el manejo del m��dulo irradiado. Por ��ltimo, en esta tesis se ha estudiado tambi��n las posibilidades existentes para la fabricaci��n del m��dulo. De entre las t��cnicas propuestas, destacan la electroerosi��n, soldaduras por haz de electrones o por soldadura l��ser. Las bases para el dise��o final del LBVM han sido pues establecidas en el marco de este trabajo y han sido incluidas en el dise��o intermedio de IFMIF, que ser�� desarrollado en el futuro, como parte del dise��o final de la instalaci��n IFMIF. ABSTRACT Nuclear fusion is, today, an alternative energy source to which the international community devotes a great effort. The goal is to generate 10 to 50 times more energy than the input power by means of fusion reactions that occur in deuterium (D) and tritium (T) plasma at two hundred million degrees Celsius. In the future commercial reactors it will be necessary to breed the tritium used as fuel in situ, by the reactor itself. This constitutes a step further from current experimental machines dedicated mainly to the study of the plasma physics. Therefore, tritium, in fusion reactors, will be produced in the so-called breeder blankets whose primary mission is to provide neutron shielding, produce and recover tritium and convert the neutron energy into heat. There are different concepts of breeding blankets that can be separated into two main categories: solids or liquids. The former are based on ceramics containing lithium as Li2O , Li4SiO4 , Li2TiO3 , Li2ZrO3 and Be, used as a neutron multiplier, required to achieve the required amount of tritium. The liquid concepts are based on molten salts or liquid metals as pure Li, LiPb, FLIBE or FLINABE. These blankets use, as neutron multipliers, Be or Pb (in the case of the concepts based on LiPb). Proposed structural materials comprise various options, always with low activation characteristics, as low activation ferritic-martensitic steels, vanadium alloys or even SiCf/SiC. Each concept of breeding blanket has specific challenges that will be studied in the experimental reactor ITER (International Thermonuclear Experimental Reactor). However, ITER cannot answer questions associated to material damage and the effect of neutron radiation in the different breeding blankets functions and performance. As a reference, the first wall of a fusion reactor of 4000 MW will receive about 30 dpa / year (values for Fe-56) , while values expected in ITER would be <10 dpa in its entire lifetime. Consequently, the irradiation effects on candidate materials for fusion reactors will be studied in IFMIF (International Fusion Material Irradiation Facility). This thesis fits in the framework of the bilateral agreement among Europe and Japan which is called ���Broader Approach Agreement ���(BA) (2007-2017) where Spain plays a key role. These projects, complementary to ITER, are mainly IFMIF and the fusion facility JT-60SA. The purpose of this thesis is the design of an irradiation module to test candidate materials for breeding blankets in IFMIF, the so-called Liquid Breeder Validation Module (LBVM). This proposal is born from the fact that this option was not considered in the conceptual design of the facility. As a first step, in order to study the feasibility of this proposal, neutronic calculations have been performed to estimate irradiation parameters in different materials foreseen for liquid breeding blankets. Various functional materials were considered: Fe (base of structural materials), SiC (candidate material for flow channel inserts, SiO2 (candidate for antipermeation coatings), CaO (candidate for insulating coatings), Al2O3 (candidate for antipermeation and insulating coatings) and AlN (candidate for insulation coating material). For each material, the most significant irradiation parameters have been calculated (dpa, H/dpa and He/dpa) in different positions of IFMIF. These values were compared to those expected in the first wall and breeding zone of a fusion reactor. For this exercise, a HCLL (Helium Cooled Lithium Lead) type was selected as it is one of the most promising options. In addition, estimated values were also compared with those obtained in a fast fission reactor since most of existing irradiations have been made in these installations. The main conclusion of this study is that the medium flux area of IFMIF offers a good irradiation environment to irradiate functional materials for liquid breeding blankets. The obtained ratios of H/dpa and He/dpa are very similar to those expected in the most irradiated areas of a fusion reactor. Moreover, with the aim of bringing the values further close, the use of a W moderator is proposed to be used only in some experimental campaigns (as obviously, the total amount of dpa decreases). The values of ratios obtained for a fission reactor, much lower than in a fusion reactor, reinforce the need of LBVM for IFMIF. Having demonstrated the suitability of IFMIF to irradiate functional materials for liquid breeding blankets, and an analysis of the main problems associated to each type of liquid breeding blanket, also presented in this thesis, three different experiments are proposed as basis for the design of the LBVM. These experiments are dedicated to the needs of a blanket HCLL type although the applicability of the module for other blankets is also discussed. Therefore, the experimental capability of the module is focused on the study of the behavior of the eutectic alloy LiPb, tritium permeation, corrosion and material compatibility. For each of the experiments proposed an experimental scheme is given explaining the different module conditions and defining the required instrumentation to control and monitor the experimental capsules. In order to carry out the proposed experiments, the LBVM is proposed, located in the medium flux area of the IFMIF hot cell, with capability of up to 16 experimental capsules. Each capsule (24-22 mm of diameter, 80 mm high) will contain the eutectic allow LiPb (up to 50 mm of capsule high) in contact with different material specimens. They will be supported inside rigs or steel pipes. Helium will be used as purge gas, to sweep the tritium generated in the eutectic and permeated through the capsule walls (continuously, during irradiation). These tubes, will be installed in a steel container providing support and cooling for the tubes and hence the inner experimental capsules. The experimental data will consist of on line monitoring signals and the analysis of purge gas by the tritium measurement station. In addition to the experimental signals, the module will produce signals having a safety function and therefore playing a major role in the operation of the module. For an adequate operation of the capsules and to control its temperature, each capsule will be equipped with an electrical heater so the module will to be connected to an electrical power supply. The technical justification behind the dimensioning of each of these parts forming the module is presented supported by tritium transport calculations, thermalhydraulic and structural analysis. One of the main conclusions of the tritium transport calculations is that the measure of the permeated tritium is perfectly achievable by commercial ionization chambers and proportional counters with sensitivity of 10-9 Bq/m3. The results are applicable to all experiments, even to low temperature capsules or to the ones using antipermeation coatings. From a safety point of view, the knowledge of the amount of tritium being swept by the purge gas is a clear indicator of certain problems that may be occurring in the module such a capsule rupture. In addition, the tritium balance in the installation should be known. Losses of purge gas permeated into the refrigerant and the hot cell itself through the container have been assessed concluding that they are negligible for normal operation. Thermal hydraulic calculations were performed in order to optimize the design of experimental capsules and LBVM to fulfill one of the main requirements of the module: to perform experiments at uniform temperatures between 300-550��C. The necessary cooling of the module and the geometry of the capsules, rigs and testing area of the container were dimensioned. As a result of the analyses, cylindrical capsules and rigs in cylindrical compartments were selected because of their good mechanical behavior (stresses due to fluid pressure are reduced significantly with a cylindrical shape rather than prismatic) and thermal (temperature uniformity in the walls of the tubes and capsules). Fields of pressure, temperature and velocity in different critical areas of the module were obtained concluding that the proposal is feasible. It is important to mention that the use of fluid dynamic codes as ANSYS-CFX (used in this thesis) for designing experimental capsules for IFMIF is not direct. The reason for this is that, under strongly heated helium mini channels, turbulence models tend to underestimate the wall temperature because of the change of helium properties near the wall. Therefore, the different code turbulence models had to be studied in detail and validated against experimental results. ANSYS-CFX SST (Shear Stress Transport Model) for transitional turbulence model has been identified among many others as the suitable one for modeling the cooling helium and the temperature on the walls of experimental capsules. Once the geometry and the main purge and cooling parameters have been defined, the mechanical behavior of each experimental tube or rig including capsules is analyzed. Resulting stresses are compared with the maximum values recommended by applicable structural design codes such as the SDC- IC (Structural Design Criteria for ITER Components) in order to assess the degree of protection against plastic collapse. The conclusion shows that the proposal is mechanically robust. The LBVM involves the use of liquid metals, tritium and the risk associated with neutron activation. The risks related with the handling of liquid metals and tritium are studied in this thesis. In addition, the radiological risks associated with the activation of materials in the module and the residual heat after irradiation are evaluated, including a scenario of loss of coolant. Among the identified conventional risks associated with the module highlights the handling of liquid metals which reactions with water or air are accompanied by the emission of aerosols and fire probability. Regarding the nuclear risks, the generation of radioactive gases such as tritium or volatile radioisotopes such as Po-210 is the main hazard to be considered. An environmental impact associated to possible releases is not expected. Nevertheless, an appropriate handling of capsules, experimental tubes, and container including purge lines is required. After one day after shutdown and one year of irradiation, the experimental area of the module will present a contact dose rate of about 7000 Sv/h, 2300 Sv/h in the experimental capsules and 25 Sv/h in the LiPb. Therefore, the use of remote handling is envisaged for the irradiated module. Finally, the different possibilities for the module manufacturing have been studied. Among the proposed techniques highlights the electro discharge machining, brazing, electron beam welding or laser welding. The bases for the final design of the LBVM have been included in the framework of the this work and included in the intermediate design report of IFMIF which will be developed in future, as part of the IFMIF facility final design.

Antigen-driven CD4+ T cell and HIV-1 dynamics: Residual viral replication under highly active antiretroviral therapy

Antigen-induced stimulation of the immune system can generate heterogeneity in CD4+ T cell division rates capable of explaining the temporal patterns seen in the decay of HIV-1 plasma RNA levels during highly active antiretroviral therapy. Posttreatment increases in peripheral CD4+ T cell counts are consistent with a mathematical model in which host cell redistribution between lymph nodes and peripheral blood is a function of viral burden. Model fits to patient data suggest that, although therapy reduces HIV replication below replacement levels, substantial residual replication continues. This residual replication has important consequences for long-term therapy and the evolution of drug resistance and represents a challenge for future treatment strategies.

Instabilidades MHD no Tokamak TCABR

Este trabalho descreve o estudo das instabilidades magneto-hidro-din��micas (MHD) comumente observadas nas descargas el��tricas de plasma no tokamak TCABR, do Instituto de F��sica da USP. Dois diagn��sticos principais foram empregados para observar essas instabilidades: um conjunto poloidal de 24 bobinas magn��ticas (bobinas de Mirnov) colocadas pr��ximas �� borda do plasma e um medidor de emiss��es na faixa do Ultra Violeta e de raios X moles com 20 canais (sistema SXR), cujo circuito de condicionamento de sinais foi aprimorado como parte deste trabalho. Esses diagn��sticos foram escolhidos porque fornecem informa����es complementares, uma vez que o sistema SXR observa a parte central da coluna de plasma, enquanto as bobinas de Mirnov detectam as instabilidades MHD na regi��o mais externa da coluna. As informa����es coletadas por esses diagn��sticos foram submetidas �� an��lise espectral com resolu����o temporal e espacial, possibilitando determinar a evolu����o das caracter��sticas espectrais e espaciais das instabilidades MHD observadas. Essas an��lises revelaram que durante a etapa inicial da forma����o do plasma (quando a corrente de plasma ainda est�� aumentando) ilhas magn��ticas com n��meros de onda decrescente, identificadas como sendo modos kink de borda, s��o detectadas nas bobinas de Mirnov. Ap��s a forma����o do plasma, quando os par��metros de equil��brio est��o relativamente est��veis (plat��), oscila����es s��o detectadas tanto nas bobinas de Mirnov quanto no sistema de SXR, indicando a presen��a de instabilidades MHD em toda a coluna de plasma. Em geral as oscila����es medidas nas bobinas de Mirnov tem baixa amplitude e correspondem a pequenas ilhas magn��ticas que foram identificadas como sendo modos de ruptura (modos tearing). Por outro lado, as instabilidades na regi��o central foram identificadas como dentes de serra, que correspondem a relaxa����es peri��dicas da regi��o interna �� superf��cie magn��tica com fator de seguran��a q=1 e que s��o acompanhadas de oscila����es precursoras, cuja amplitude depende da fase do ciclo de relaxa����o. Devido �� essa modula����o de amplitude, aparecem picos de frequ��ncia sat��lite nos espectrogramas dos sinais do SXR. Al��m disso, devido ao fato dos ciclos de relaxa����o n��o serem sinusoidais, os harm��nicos da frequ��ncia de relaxa����o tamb��m aparecem nesses espectrogramas. No entanto, em muitas descargas do TCABR, a intensidade das oscila����es medidas nas bobinas de Mirnov aumentam significativamente durante o plat��, com efeitos sobre a frequ��ncia de todas as instabilidades MHD, at�� mesmo sobre os dentes de serra localizados na regi��o central da coluna. Em todos os casos, observou-se que durante o plat�� a frequ��ncia das ilhas magn��ticas coincide com a frequ��ncia das oscila����es precursoras do dente de serra, apesar de serem duas instabilidades distintas, localizadas em posi����es radiais muito diferentes. Essa coincid��ncia de frequ��ncias possibilitou descrever a evolu����o em frequ��ncia de todas as oscila����es detectadas em diversos diagn��sticos com base em apenas duas frequ��ncias b��sicas: a dos ciclos de relaxa����o dente de serra e a das ilhas magn��ticas.

Principles of plasma dynamics.

"These notes are the result of the author's lectures [sic] on plasma physics in the spring term of 1961 at the University of Miami."

A Kinetic Vlasov Model for Plasma Simulation Using Discontinuous Galerkin Method on Many-Core Architectures

Thesis (Ph.D.)--University of Washington, 2016-03

Parametric Optimization of Inductively Driven Liner Compression Driver System

Thesis (Master's)--University of Washington, 2016-06

Scaling of multiple postselected quantum gates in optics

We show that interesting multigate circuits can be constructed using a postselected controlled-sign gate that works with a probability (1/3)(n), where n-1 is the number of controlled-sign gates in the circuit, rather than (1/9)(n-1), as would be expected from a sequence of such gates. We suggest some quantum information tasks which could be demonstrated using these circuits, such as parity checking and cluster-state computation.

Coherent-state linear optical quantum computing gates using simplified diagonal superposition resource states

In this paper we explore the possibility of fundamental tests for coherent-state optical quantum computing gates [ T. C. Ralph et al. Phys. Rev. A 68 042319 (2003)] using sophisticated but not unrealistic quantum states. The major resource required in these gates is a state diagonal to the basis states. We use the recent observation that a squeezed single-photon state [S(r)���1���] approximates well an odd superposition of coherent states (����������������������) to address the diagonal resource problem. The approximation only holds for relatively small ��, and hence these gates cannot be used in a scalable scheme. We explore the effects on fidelities and probabilities in teleportation and a rotated Hadamard gate.

High-fidelity Z-measurement error encoding of optical qubits

We demonstrate a quantum error correction scheme that protects against accidental measurement, using a parity encoding where the logical state of a single qubit is encoded into two physical qubits using a nondeterministic photonic controlled-NOT gate. For the single qubit input states vertical bar 0 >, vertical bar 1 >, vertical bar 0 > +/- vertical bar 1 >, and vertical bar 0 > +/- i vertical bar 1 > our encoder produces the appropriate two-qubit encoded state with an average fidelity of 0.88 +/- 0.03 and the single qubit decoded states have an average fidelity of 0.93 +/- 0.05 with the original state. We are able to decode the two-qubit state (up to a bit flip) by performing a measurement on one of the qubits in the logical basis; we find that the 64 one-qubit decoded states arising from 16 real and imaginary single-qubit superposition inputs have an average fidelity of 0.96 +/- 0.03.