29 resultados para . neutron radiation field
em Universidad Politécnica de Madrid
Resumo:
This work discusses an iterative procedure of shaping offset dual-reflector antennas based on geometrical optics considering both far-field and near-field measurements of amplitude and phase from the feed horn. The surfaces synthesized will transform a known radiation field of a feed to a desired aperture distribution. This technique is applied for both circular and elliptical apertures and has the advantage to simplify the problem compared with existing techniques based on solving nonlinear differential equations. A MATLAB tool has been developed to implement the shaping algorithms. This procedure is applied for the design of a 1.1 m high-gain antenna for the ESAs Solar Orbiter spacecraft. This antenna operating at X-band will manage high data rate and high efficiency communications with Earth stations.
Resumo:
La fusin nuclear es, hoy en da, una alternativa energtica a la que la comunidad internacional dedica mucho esfuerzo. El objetivo es el de generar entre diez y cincuenta veces ms energa que la que consume mediante reacciones de fusin que se producirn en una mezcla de deuterio (D) y tritio (T) en forma de plasma a doscientos millones de grados centgrados. En los futuros reactores nucleares de fusin ser necesario producir el tritio utilizado como combustible en el propio reactor termonuclear. Este hecho supone dar un paso ms que las actuales mquinas experimentales dedicadas fundamentalmente al estudio de la fsica del plasma. As pues, el tritio, en un reactor de fusin, se produce en sus envolturas regeneradoras cuya misin fundamental es la de blindaje neutrnico, producir y recuperar tritio (fuel para la reaccin DT del plasma) y por ltimo convertir la energa de los neutrones en calor. Existen diferentes conceptos de envolturas que pueden ser slidas o lquidas. Las primeras se basan en cermicas de litio (Li2O, Li4SiO4, Li2TiO3, Li2ZrO3) y multiplicadores neutrnicos de Be, necesarios para conseguir la cantidad adecuada de tritio. Los segundos se basan en el uso de metales lquidos o sales fundidas (Li, LiPb, FLIBE, FLINABE) con multiplicadores neutrnicos de Be o el propio Pb en el caso de LiPb. Los materiales estructurales pasan por aceros ferrtico-martensticos de baja activacin, aleaciones de vanadio o incluso SiCf/SiC. Cada uno de los diferentes conceptos de envoltura tendr una problemtica asociada que se estudiar en el reactor experimental ITER (del ingls, International Thermonuclear Experimental Reactor). Sin embargo, ITER no puede responder las cuestiones asociadas al dao de materiales y el efecto de la radiacin neutrnica en las diferentes funciones de las envolturas regeneradoras. Como referencia, la primera pared de un reactor de fusin de 4000MW recibira 30 dpa/ao (valores para Fe-56) mientras que en ITER se conseguiran <10 dpa en toda su vida til. Esta tesis se encuadra en el acuerdo bilateral entre Europa y Japn denominado Broader Approach Agreement (BA) (2007-2017) en el cual Espaa juega un papel destacable. Estos proyectos, complementarios con ITER, son el acelerador para pruebas de materiales IFMIF (del ingls, International Fusion Materials Irradiation Facility) y el dispositivo de fusin JT-60SA. As, los efectos de la irradiacin de materiales en materiales candidatos para reactores de fusin se estudiarn en IFMIF. El objetivo de esta tesis es el diseo de un mdulo de IFMIF para irradiacin de envolturas regeneradoras basadas en metales lquidos para reactores de fusin. El mdulo se llamar LBVM (del ingls, Liquid Breeder Validation Module). La propuesta surge de la necesidad de irradiar materiales funcionales para envolturas regeneradoras lquidas para reactores de fusin debido a que el diseo conceptual de IFMIF no contaba con esta utilidad. Con objeto de analizar la viabilidad de la presente propuesta, se han realizado clculos neutrnicos para evaluar la idoneidad de llevar a cabo experimentos relacionados con envolturas lquidas 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 FCIs (del ingls, Flow Channel Inserts) en una envoltura regeneradora lquida, SiO2 (candidato para recubrimientos antipermeacin), CaO (candidato para recubrimientos aislantes), Al2O3 (candidato para recubrimientos antipermeacin y aislantes) y AlN (material candidato para recubrimientos aislantes). En cada uno de estos materiales se han calculado los parmetros de irradiacin ms 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 fusin. Para ello se ha elegido un reactor tipo HCLL (del ingls, Helium Cooled Lithium Lead) por tratarse de uno de los ms prometedores. Adems, los valores tambin se han comparado con los que se obtendran en un reactor rpido de fisin puesto que la mayora de las irradiaciones actuales se hacen en reactores de este tipo. Como conclusin al anlisis de viabilidad, se puede decir que los materiales funcionales para mantos regeneradores lquidos podran probarse en la zona de medio flujo de IFMIF donde se obtendran ratios de H/dpa y He/dpa muy parecidos a los esperados en las zonas ms irradiadas de un reactor de fusin. Adems, con el objetivo de ajustar todava ms los valores, se propone el uso de un moderador de W (a considerar en algunas campaas de irradiacin solamente debido a que su uso hace que los valores de dpa totales disminuyan). Los valores obtenidos para un reactor de fisin 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 fusin y, por lo tanto, no representativos. Una vez demostrada la idoneidad de IFMIF para irradiar envolturas regeneradoras lquidas, y del estudio de la problemtica asociada a las envolturas lquidas, tambin incluida en esta tesis, se proponen tres tipos de experimentos diferentes como base de diseo 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 mdulo estara centrada en el estudio del comportamiento de litio plomo, permeacin de tritio, corrosin y compatibilidad de materiales. Para cada uno de los experimentos se propone un esquema experimental, se definen las condiciones necesarias en el mdulo y la instrumentacin requerida para controlar y diagnosticar las cpsulas 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 cpsulas experimentales. Cada cpsula (24-22 mm de dimetro y 80 mm de altura) contendr la aleacin eutctica LiPb (hasta 50 mm de la altura de la cpsula) 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 eutctico y permeado a travs de las paredes de las cpsulas (continuamente, durante irradiacin). Estos tubos, a su vez, se instalarn en una carcasa tambin de acero que proporcionar soporte y refrigeracin tanto a los tubos como a sus cpsulas experimentales interiores. El mdulo, en su conjunto, permitir la extraccin de las seales experimentales y el gas de purga. As, a travs de la estacin de medida de tritio y el sistema de control, se obtendrn los datos experimentales para su anlisis y extraccin de conclusiones experimentales. Adems del anlisis de datos experimentales, algunas de estas seales tendrn una funcin de seguridad y por tanto jugarn un papel primordial en la operacin del mdulo. Para el correcto funcionamiento de las cpsulas y poder controlar su temperatura, cada cpsula se equipar con un calentador elctrico y por tanto el mdulo requerir tambin ser conectado a la alimentacin elctrica. El diseo del mdulo y su lgica de operacin se describe en detalle en esta tesis. La justificacin tcnica de cada una de las partes que componen el mdulo se ha realizado con soporte de clculos de transporte de tritio, termohidrulicos y mecnicos. Una de las principales conclusiones de los clculos de transporte de tritio es que es perfectamente viable medir el tritio permeado en las cpsulas mediante cmaras de ionizacin 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 cpsulas a bajas temperaturas o si llevan recubrimientos antipermeacin. 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 mdulo como por ejemplo, la rotura de una cpsula. Adems, es necesario conocer el balance de tritio de la instalacin. Las prdidas esperadas el refrigerante y la celda caliente de IFMIF se pueden considerar despreciables para condiciones normales de funcionamiento. Los clculos termohidrulicos se han realizado con el objetivo de optimizar el diseo de las cpsulas experimentales y el LBVM de manera que se pueda cumplir el principal requisito del mdulo que es llevar a cabo los experimentos a temperaturas comprendidas entre 300-550C. Para ello, se ha dimensionado la refrigeracin necesaria del mdulo y evaluado la geometra de las cpsulas, tubos experimentales y la zona experimental del contenedor. Como consecuencia de los anlisis realizados, se han elegido cpsulas y tubos cilndricos instalados en compartimentos cilndricos debido a su buen comportamiento mecnico (las tensiones debidas a la presin de los fluidos se ven reducidas significativamente con una geometra cilndrica en lugar de prismtica) y trmico (uniformidad de temperatura en las paredes de los tubos y cpsulas). Se han obtenido campos de presin, temperatura y velocidad en diferentes zonas crticas del mdulo concluyendo que la presente propuesta es factible. Cabe destacar que el uso de cdigos fluidodinmicos (e.g. ANSYS-CFX, utilizado en esta tesis) para el diseo de cpsulas experimentales de IFMIF no es directo. La razn 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 cdigo han tenido que ser estudiados con detalle y validados con resultados experimentales. El modelo SST (del ingls, Shear Stress Transport Model) para turbulencia en transicin ha sido identificado como adecuado para simular el comportamiento del helio de refrigeracin y la temperatura en las paredes de las cpsulas experimentales. Con la geometra propuesta y los valores principales de refrigeracin y purga definidos, se ha analizado el comportamiento mecnico de cada uno de los tubos experimentales que contendr el mdulo. Los resultados de tensiones obtenidos, han sido comparados con los valores mximos recomendados en cdigos de diseo estructural como el SDC-IC (del ingls, Structural Design Criteria for ITER Components) para as evaluar el grado de proteccin contra el colapso plstico. La conclusin del estudio muestra que la propuesta es mecnicamente robusta. El LBVM implica el uso de metales lquidos y la generacin de tritio adems del riesgo asociado a la activacin neutrnica. Por ello, se han estudiado los riesgos asociados al uso de metales lquidos y el tritio. Adems, se ha incluido una evaluacin preliminar de los riesgos radiolgicos asociados a la activacin de materiales y el calor residual en el mdulo despus de la irradiacin as como un escenario de prdida de refrigerante. Los riesgos asociados al mdulo de naturaleza convencional estn asociados al manejo de metales lquidos cuyas reacciones con aire o agua se asocian con emisin de aerosoles y probabilidad de fuego. De entre los riesgos nucleares destacan la generacin de gases radiactivos como el tritio u otros radioistopos voltiles como el Po-210. No se espera que el mdulo suponga un impacto medioambiental asociado a posibles escapes. Sin embargo, es necesario un manejo adecuado tanto de las cpsulas experimentales como del mdulo contenedor as como de las lneas de purga durante operacin. Despus de un da de despus de la parada, tras un ao de irradiacin, tendremos una dosis de contacto de 7000 Sv/h en la zona experimental del contenedor, 2300 Sv/h en la cpsula y 25 Sv/h en el LiPb. El uso por lo tanto de manipulacin remota est previsto para el manejo del mdulo irradiado. Por ltimo, en esta tesis se ha estudiado tambin las posibilidades existentes para la fabricacin del mdulo. De entre las tcnicas propuestas, destacan la electroerosin, soldaduras por haz de electrones o por soldadura lser. Las bases para el diseo final del LBVM han sido pues establecidas en el marco de este trabajo y han sido incluidas en el diseo intermedio de IFMIF, que ser desarrollado en el futuro, como parte del diseo final de la instalacin 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 300-550C. 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.
Resumo:
Monte Carlo calculations were carried out to characterize the neutron field produced by the calibration neutron sources of the Neutron Standards Laboratory at the Research Center for Energy, Environment and Technology (CIEMAT) in Spain. For 241AmBe and 252Cf neutron sources, the neutron spectra, the ambient dose equivalent rates and the total neutron fluence rates were estimated. In the calibration hall, there are several items that modify the neutron field. To evaluate their effects different cases were used, from point-like source in vacuum up to the full model. Additionally, using the full model, the neutron spectra were estimated to different distances along the bench; with these spectra, the total neutron fluence and the ambient dose equivalent rates were calculated. The hall walls induce the largest changes in the neutron spectra and the respective integral quantities. The free-field neutron spectrum is modified due the room return effect.
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The engineering design of fissionchambers as on-line radiation detectors for IFMIF is being performed in the framework of the IFMIF-EVEDA works. In this paper the results of the experiments performed in the BR2 reactor during the phase-2 of the foreseen validation activities are addressed. Two detectors have been tested in a mixedneutron-gamma field with high neutron fluence and gamma absorbed dose rates, comparable with the expected values in the HFTM in IFMIF. Since the neutron spectra in all BR2 channels are dominated by the thermal neutron component, the detectors have been surrounded by a cylindrical gadolinium screen to cut the thermal neutron component, in order to get a more representative test for IFMIF conditions. The integrated gamma absorbed dose was about 4 1010 Gy and the fast neutron fluence (E > 0.1 MeV) 4 1020 n/cm2. The fissionchambers were calibrated in three BR2 channels with different neutron-to-gamma ratio, and the long-term evolution of the signals was studied and compared with theoretical calculations
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A new and effective method for reduction of truncation errors in partial spherical near-field (SNF) measurements is proposed. The method is useful when measuring electrically large antennas, where the measurement time with the classical SNF technique is prohibitively long and an acquisition over the whole spherical surface is not practical. Therefore, to reduce the data acquisition time, partial sphere measurement is usually made, taking samples over a portion of the spherical surface in the direction of the main beam. But in this case, the radiation pattern is not known outside the measured angular sector as well as a truncation error is present in the calculated far-field pattern within this sector. The method is based on the Gerchberg-Papoulis algorithm used to extrapolate functions and it is able to extend the valid region of the calculated far-field pattern up to the whole forward hemisphere. To verify the effectiveness of the method, several examples are presented using both simulated and measured truncated near-field data.
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This paper describes two methods to cancel the effect of two kinds of leakage signals which may be presented when an antenna is measured in a planar near-field range. One method tries to reduce leakage bias errors from the receivers quadrature detector and it is based on estimating the bias constant added to every near-field data sample. Then, that constant is subtracted from the data, removing its undesired effect on the far-field pattern. The estimation is performed by back-propagating the field from the scan plane to the antenna under test plane (AUT) and averaging all the data located outside the AUT aperture. The second method is able to cancel the effect of the leakage from faulty transmission lines, connectors or rotary joints. The basis of this method is also a reconstruction process to determine the field distribution on the AUT plane. Once this distribution is known, a spatial filtering is applied to cancel the contribution due to those faulty elements. After that, a near-field-to-far-field transformation is applied, obtaining a new radiation pattern where the leakage effects have disappeared. To verify the effectiveness of both methods, several examples are presented.
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Neutron spectra unfolding and dose equivalent calculation are complicated tasks in radiation protection, are highly dependent of the neutron energy, and a precise knowledge on neutron spectrometry is essential for all dosimetry-related studies as well as many nuclear physics experiments. In previous works have been reported neutron spectrometry and dosimetry results, by using the ANN technology as alternative solution, starting from the count rates of a Bonner spheres system with a LiI(Eu) thermal neutrons detector, 7 polyethylene spheres and the UTA4 response matrix with 31 energy bins. In this work, an ANN was designed and optimized by using the RDANN methodology for the Bonner spheres system used at CIEMAT Spain, which is composed of a He neutron detector, 12 moderator spheres and a response matrix for 72 energy bins. For the ANN design process a neutrons spectra catalogue compiled by the IAEA was used. From this compilation, the neutrons spectra were converted from lethargy to energy spectra. Then, the resulting energy ?uence spectra were re-binned by using the MCNP code to the corresponding energy bins of the He response matrix before mentioned. With the response matrix and the re-binned spectra the counts rate of the Bonner spheres system were calculated and the resulting re-binned neutrons spectra and calculated counts rate were used as the ANN training data set.
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The faade is the visible part of a building, and generally consists of various different constructive systems. The sound reduction index of the closing elements for the openings on a rooms faade is a determining factor in the sound insulation from airborne noise inside the space. Windows are the transparent part of the faade, and to improve their thermal behaviour and control solar radiation, they are often fitted with a series of external and internal protections such as shutters, slats and blinds. This work contains a summary of studies carried out using field measurements of airborne sound insulation on faades in rooms, in application of the standard UNE-EN ISO 140-5:1999. In all the rooms the windows were fitted with shutter boxes and rolling shutters, and the acoustic tests were made with the shutter in two positions (extended and fully retracted). The results were analysed considering the window opening system (openable or sliding) and the type of glass pane (monolithic or insulating glass unit, IGU). In the case of sliding windows, the airborne sound insulation of faades is greater when the shutter is extended than when it is retracted, and this should be taken into account when applying the aforementioned standard.
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The neutronics hall of the Nuclear Engineering Department at the Polytechnical University of Madrid has been characterized. The neutron spectra and the ambient dose equivalent produced by an 241AmBe source were measured at various source-to-detector distances on the new bench. Using Monte Carlo methods a detailed model of the neutronics hall was designed, and neutron spectra and the ambient dose equivalent were calculated at the same locations where measurements were carried out. A good agreement between measured and calculated values was found.
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In this paper, a fully automatic goal-oriented hp-adaptive finite element strategy for open region electromagnetic problems (radiation and scattering) is presented. The methodology leads to exponential rates of convergence in terms of an upper bound of an user-prescribed quantity of interest. Thus, the adaptivity may be guided to provide an optimal error, not globally for the field in the whole finite element domain, but for specific parameters of engineering interest. For instance, the error on the numerical computation of the S-parameters of an antenna array, the field radiated by an antenna, or the Radar Cross Section on given directions, can be minimized. The efficiency of the approach is illustrated with several numerical simulations with two dimensional problem domains. Results include the comparison with the previously developed energy-norm based hp-adaptivity.
Resumo:
An automated panoramic irradiator with a 3 Ci 241Am-Be neutron source is installed in a bunker-type large room at the Universidad Politcnica de Madrid (UPM). It was recently modified and a neutron spectrometry campaign was organized to characterize the neutron fields in different measurement points along the irradiation bench. Four research groups working with different Bonner Sphere Spectrometers (BSS) and using different spectral unfolding codes took part to this exercise. INFN-LNF used a BSS formed by 9 spheres plus bare detector, with cylindrical, almost point like, 6LiI(Eu) scintillator (4 mm x 4 mm, from Ludlum); UAZ-UPM employed a similar system but with only 6 spheres plus bare detector; UAB worked with a 3He filled proportional counter at 8kPa filling pressure, cylindrical 9 mm x 10 mm (05NH1 from Eurisys) with 11 spheres configuration; and CIEMAT used 12 spheres with an spherical 3He SP9 counter (Centronic Ltd., UK) with very high sensitivity due to the large diameter (3.2 cm) and the filling pressure of the order of 228 kPa. Each group applied a different spectral unfolding method: INFN and UAB worked with FRUIT ver. 3.0 with their own response matrixes; UAZ-UPM used the BUNKIUT unfolding code with the response matrix UTA4 and CIEMAT employed the GRAVEL-MAXED-IQU package with their own response matrix. The paper shows the main results obtained in terms of neutron spectra at fixed distances from the source as well as total neutron fluence rate and ambient dose equivalent rate H*(10) determined from the spectra. The latter are compared with the readings of a common active survey-meter (LB 6411). The small differences in the results of the various groups are discussed.
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Kinetic Monte Carlo (KMC) is a widely used technique to simulate the evolution of radiation damage inside solids. Despite de fact that this technique was developed several decades ago, there is not an established and easy to access simulating tool for researchers interested in this field, unlike in the case of molecular dynamics or density functional theory calculations. In fact, scientists must develop their own tools or use unmaintained ones in order to perform these types of simulations. To fulfil this need, we have developed MMonCa, the Modular Monte Carlo simulator. MMonCa has been developed using professional C++ programming techniques and has been built on top of an interpreted language to allow having a powerful yet flexible, robust but customizable and easy to access modern simulator. Both non lattice and Lattice KMC modules have been developed. We will present in this conference, for the first time, the MMonCa simulator. Along with other (more detailed) contributions in this meeting, the versatility of MMonCa to study a number of problems in different materials (particularly, Fe and W) subject to a wide range of conditions will be shown. Regarding KMC simulations, we have studied neutron-generated cascade evolution in Fe (as a model material). Starting with a Frenkel pair distribution we have followed the defect evolution up to 450 K. Comparison with previous simulations and experiments shows excellent agreement. Furthermore, we have studied a more complex system (He-irradiated W:C) using a previous parametrization [1]. He-irradiation at 4 K followed by isochronal annealing steps up to 500 K has been simulated with MMonCa. The He energy was 400 eV or 3 keV. In the first case, no damage is associated to the He implantation, whereas in the second one, a significant Frenkel pair concentration (evolving into complex clusters) is associated to the He ions. We have been able to explain He desorption both in the absence and in the presence of Frenkel pairs and we have also applied MMonCa to high He doses and fluxes at elevated temperatures. He migration and trapping dominate the kinetics of He desorption. These processes will be discussed and compared to experimental results. [1] C.S. Becquart et al. J. Nucl. Mater. 403 (2010) 75
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The dispersion relation for waves in a cold, magnetized plasma is discussed using the potential for the longitudinal part of the electric field. This clarifies wave emission from a conductor in low Earth orbit and should be useful in considering the far field and both hot plasma and nonlinear, near-field effects. General formulas for radiation impedance are directly obtained. For tethers a fundamental dependence on contactor size is discussed. Spherical and ellipsoidal contactors and an (anodcless) bare tether are considered. Simple arguments on nonlinear contactor effects lead to a surprisingly simple result for impedances off the Alfven branch.
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Juno, the second mission in the NASA New Frontiers Program, will both be a polar Jovian orbiter, and use solar arrays for power, moving away from previous use of radioisotope power systems (RPSs) in spite of the weak solar light reaching Jupiter. The power generation at Jupiter is critical, and a conductive tether could be an alternative source of power. A current-carrying tether orbiting in a magnetized ionosphere/plasmasphere will radiate waves. A magnitude of interest for both power generation and signal emission is the wave impedance. Jupiter has the strongest magnetic field in the Solar Planetary System and its plasma density is low everywhere. This leads to an electron plasma frequency smaller than the electron cyclotron frequency, and a high Alfven velocity. Unlike the low Earth orbit (LEO) case, the electron skin depth and the characteristic size of plasma contactors affect the Alfven impedance.
Resumo:
Wave radiation by a conductor carrying a steady current in both a polar, highly eccentric, low perijove orbit, as in NASA's planned Juno mission, and an equatorial low Jovian orbit (LJO) mission below the intense radiation belts, is considered. Both missions will need electric power generation for scientific instruments and communication systems. Tethers generate power more efficiently than solar panels or radioisotope power systems (RPS). The radiation impedance is required to determine the current in the overall tether circuit. In a cold plasma model, radiation occurs mainly in the Alfven and fast magnetosonic modes, exhibiting a large refraction index. The radiation impedance of insulated tethers is determined for both modes and either mission. Unlike the Earth ionospheric case, the low-density, highly magnetized Jovian plasma makes the electron gyrofrequency much larger than the plasma frequency; this substantially modifies the power spectrum for either mode by increasing the Alfven velocity. Finally, an estimation of the radiation impedance of bare tethers is considered. In LJO, a spacecraft orbiting in a slow downward spiral under the radiation belts would allow determining magnetic field structure and atmospheric composition for understanding the formation, evolution, and structure of Jupiter. Additionally, if the cathodic contactor is switched off, a tether floats electrically, allowing e-beam emission that generate auroras. On/off switching produces bias/current pulses and signal emission, which might be used for Jovian plasma diagnostics.
