Advanced data acquisition system implementation for the ITER Neutron Diagnostic use case using EPICS and FlexRIO technology on a PXIe platform

In the framework of the ITER Control Breakdown Structure (CBS), Plant System Instrumentation & Control (I&C) defines the hardware and software required to control one or more plant systems [1]. For diagnostics, most of the complex Plant System I&C are to be delivered by ITER Domestic Agencies (DAs). As an example for the DAs, ITER Organization (IO) has developed several use cases for diagnostics Plant System I&C that fully comply with guidelines presented in the Plant Control Design Handbook (PCDH) [2]. One such use case is for neutron diagnostics, specifically the Fission Chamber (FC), which is responsible for delivering time-resolved measurements of neutron source strength and fusion power to aid in assessing the functional performance of ITER [3]. ITER will deploy four Fission Chamber units, each consisting of three individual FC detectors. Two of these detectors contain Uranium 235 for Neutron detection, while a third "dummy" detector will provide gamma and noise detection. The neutron flux from each MFC is measured by the three methods: . Counting Mode: measures the number of individual pulses and their location in the record. Pulse parameters (threshold and width) are user configurable. . Campbelling Mode (Mean Square Voltage): measures the RMS deviation in signal amplitude from its average value. .Current Mode: integrates the signal amplitude over the measurement period

Passive neutron area monitor with pairs of TLDs as neutron detector

A passive neutron area monitor has been designed using Monte Carlo methods; the monitor is a polyethylene cylinder with pairs of thermoluminescent dosimeters (TLD600 and TLD700) as thermal neutron detector. The monitor was calibrated with a bare and a thermalzed 241AmBe neutron sources and its performance was evaluated measuring the ambient dose equivalent due to photoneutrons produced by a 15 MV linear accelerator for radiotherapy and the neutrons in the output of a TRIGA Mark III radial beam port.

Irradiation uniformity at the Laser MegaJoule facility in the context of the shock ignition scheme

The use of the Laser MegaJoule facility within the shock ignition scheme has been considered. In the first part of the study, one-dimensional hydrodynamic calculations were performed for an inertial confinement fusion capsule in the context of the shock ignition scheme providing the energy gain and an estimation of the increase of the peak power due to the reduction of the photon penetration expected during the high-intensity spike pulse. In the second part, we considered a Laser MegaJoule configuration consisting of 176 laser beams that have been grouped providing two different irradiation schemes. In this configuration the maximum available energy and power are 1.3 MJ and 440 TW. Optimization of the laser?capsule parameters that minimize the irradiation non-uniformity during the first few ns of the foot pulse has been performed. The calculations take into account the specific elliptical laser intensity profile provided at the Laser MegaJoule and the expected beam uncertainties. A significant improvement of the illumination uniformity provided by the polar direct drive technique has been demonstrated. Three-dimensional hydrodynamic calculations have been performed in order to analyse the magnitude of the azimuthal component of the irradiation that is neglected in twodimensional hydrodynamic simulations.

Automatic fault detection on BIPV systems without solar irradiation data

BIPV systems are small PV generation units spread out over the territory, and whose characteristics are very diverse. This makes difficult a cost-effective procedure for monitoring, fault detection, performance analyses, operation and maintenance. As a result, many problems affecting BIPV systems go undetected. In order to carry out effective automatic fault detection procedures, we need a performance indicator that is reliable and that can be applied on many PV systems at a very low cost. The existing approaches for analyzing the performance of PV systems are often based on the Performance Ratio (PR), whose accuracy depends on good solar irradiation data, which in turn can be very difficult to obtain or cost-prohibitive for the BIPV owner. We present an alternative fault detection procedure based on a performance indicator that can be constructed on the sole basis of the energy production data measured at the BIPV systems. This procedure does not require the input of operating conditions data, such as solar irradiation, air temperature, or wind speed. The performance indicator, called Performance to Peers (P2P), is constructed from spatial and temporal correlations between the energy output of neighboring and similar PV systems. This method was developed from the analysis of the energy production data of approximately 10,000 BIPV systems located in Europe. The results of our procedure are illustrated on the hourly, daily and monthly data monitored during one year at one BIPV system located in the South of Belgium. Our results confirm that it is possible to carry out automatic fault detection procedures without solar irradiation data. P2P proves to be more stable than PR most of the time, and thus constitutes a more reliable performance indicator for fault detection procedures. We also discuss the main limitations of this novel methodology, and we suggest several future lines of research that seem promising to improve on these procedures.

Determination of the crystal and magnetic structure of the DyCrO4-scheelite polymorph by neutron diffraction

Neutron diffraction data of DyCrO4 oxide, prepared at 4 GPa and 833 K from the ambient pressure zircon-type, reveal that crystallize with the scheelite-type structure, space group I41/a. Accompanying this structural phase transition induced by pressure the magnetic properties change dramatically from ferromagnetism in the case of zircon to antiferromagnetism for the scheelite polymorph with a T N= 19 K. The analysis of the neutron diffraction data obtained at 1.2 K has been used to determine the magnetic structure of this DyCrO4-scheelite oxide which can be described with a k = [0, 0, 0] as propagation vector, where the Dy and Cr moments are lying in the ab-plane of the scheelite structure. The ordered magnetic moments are 10 µB and 1 µB for Dy+3 and Cr+5 respectively

Study of a-Si crystallization dependence on power and irradiation time using a CW green laser

An advantage of laser crystallization over conventional heating methods is its ability to limit rapid heating and cooling to thin surface layers. Laser energy is used to heat the a-Si thin film to change the microstructure to poly-Si. Thin film samples of a-Si were irradiated with a CW-green laser source. Laser irradiated spots were produced by using different laser powers and irradiation times. These parameters are identified as key variables in the crystallization process. The power threshold for crystallization is reduced as the irradiation time is increased. When this threshold is reached the crystalline fraction increases lineally with power for each irradiation time. The experimental results are analysed with the aid of a numerical thermal model and the presence of two crystallization mechanisms are observed: one due to melting and the other due to solid phase transformation.

Optoelectronic property analysis of MCrO4 (M=Ba, Sr) with a response to visible light irradiation

Ternary MCrO4 (M = Ba, Sr) semiconductors are materials with a variety of photocatalyst and optoelectronic applications. We present detailed microscopic analyses based on first principles of the structure, the electronic properties and the optical absorption in which the difference between symmetrically non-equivalent atoms has been considered. The high absorption coefficients of these materials are split into chemical species contributions in accordance with the symmetry. The high optical absorption in these materials is mainly because of the Cr–O inter-species transitions.

Diseño de un módulo de IFMIF para irradiación de envolturas regeneradoras basadas en metales líquidos para reactores de fusión

La fusión nuclear es, hoy en día, una alternativa energética a la que la comunidad internacional dedica mucho esfuerzo. El objetivo es el de generar entre diez y cincuenta veces más energía que la que consume mediante reacciones de fusión que se producirán en una mezcla de deuterio (D) y tritio (T) en forma de plasma a doscientos millones de grados centígrados. En los futuros reactores nucleares de fusión será necesario producir el tritio utilizado como combustible en el propio reactor termonuclear. Este hecho supone dar un paso más que las actuales máquinas experimentales dedicadas fundamentalmente al estudio de la física del plasma. Así pues, el tritio, en un reactor de fusión, se produce en sus envolturas regeneradoras cuya misión fundamental es la de blindaje neutrónico, producir y recuperar tritio (fuel para la reacción DT del plasma) y por último convertir la energía de los neutrones en calor. Existen diferentes conceptos de envolturas que pueden ser sólidas o líquidas. Las primeras se basan en cerámicas de litio (Li2O, Li4SiO4, Li2TiO3, Li2ZrO3) y multiplicadores neutrónicos de Be, necesarios para conseguir la cantidad adecuada de tritio. Los segundos se basan en el uso de metales líquidos o sales fundidas (Li, LiPb, FLIBE, FLINABE) con multiplicadores neutrónicos de Be o el propio Pb en el caso de LiPb. Los materiales estructurales pasan por aceros ferrítico-martensíticos de baja activación, aleaciones de vanadio o incluso SiCf/SiC. Cada uno de los diferentes conceptos de envoltura tendrá una problemática asociada que se estudiará en el reactor experimental ITER (del inglés, “International Thermonuclear Experimental Reactor”). Sin embargo, ITER no puede responder las cuestiones asociadas al daño de materiales y el efecto de la radiación neutrónica en las diferentes funciones de las envolturas regeneradoras. Como referencia, la primera pared de un reactor de fusión de 4000MW recibiría 30 dpa/año (valores para Fe-56) mientras que en ITER se conseguirían <10 dpa en toda su vida útil. Esta tesis se encuadra en el acuerdo bilateral entre Europa y Japón denominado “Broader Approach Agreement “(BA) (2007-2017) en el cual España juega un papel destacable. Estos proyectos, complementarios con ITER, son el acelerador para pruebas de materiales IFMIF (del inglés, “International Fusion Materials Irradiation Facility”) y el dispositivo de fusión JT-60SA. Así, los efectos de la irradiación de materiales en materiales candidatos para reactores de fusión se estudiarán en IFMIF. El objetivo de esta tesis es el diseño de un módulo de IFMIF para irradiación de envolturas regeneradoras basadas en metales líquidos para reactores de fusión. El módulo se llamará LBVM (del inglés, “Liquid Breeder Validation Module”). La propuesta surge de la necesidad de irradiar materiales funcionales para envolturas regeneradoras líquidas para reactores de fusión debido a que el diseño conceptual de IFMIF no contaba con esta utilidad. Con objeto de analizar la viabilidad de la presente propuesta, se han realizado cálculos neutrónicos para evaluar la idoneidad de llevar a cabo experimentos relacionados con envolturas líquidas en IFMIF. Así, 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 inglés, “Flow Channel Inserts”) en una envoltura regeneradora líquida, SiO2 (candidato para recubrimientos antipermeación), CaO (candidato para recubrimientos aislantes), Al2O3 (candidato para recubrimientos antipermeación y aislantes) y AlN (material candidato para recubrimientos aislantes). En cada uno de estos materiales se han calculado los parámetros de irradiación más 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 fusión. Para ello se ha elegido un reactor tipo HCLL (del inglés, “Helium Cooled Lithium Lead”) por tratarse de uno de los más prometedores. Además, los valores también se han comparado con los que se obtendrían en un reactor rápido de fisión puesto que la mayoría de las irradiaciones actuales se hacen en reactores de este tipo. Como conclusión al análisis de viabilidad, se puede decir que los materiales funcionales para mantos regeneradores líquidos podrían probarse en la zona de medio flujo de IFMIF donde se obtendrían ratios de H/dpa y He/dpa muy parecidos a los esperados en las zonas más irradiadas de un reactor de fusión. Además, con el objetivo de ajustar todavía más los valores, se propone el uso de un moderador de W (a considerar en algunas campañas de irradiación solamente debido a que su uso hace que los valores de dpa totales disminuyan). Los valores obtenidos para un reactor de fisión 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 fusión y, por lo tanto, no representativos. Una vez demostrada la idoneidad de IFMIF para irradiar envolturas regeneradoras líquidas, y del estudio de la problemática asociada a las envolturas líquidas, también incluida en esta tesis, se proponen tres tipos de experimentos diferentes como base de diseño del LBVM. Éstos 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. Así, la capacidad experimental del módulo estaría centrada en el estudio del comportamiento de litio plomo, permeación de tritio, corrosión y compatibilidad de materiales. Para cada uno de los experimentos se propone un esquema experimental, se definen las condiciones necesarias en el módulo y la instrumentación requerida para controlar y diagnosticar las cápsulas 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 cápsulas experimentales. Cada cápsula (24-22 mm de diámetro y 80 mm de altura) contendrá la aleación eutéctica LiPb (hasta 50 mm de la altura de la cápsula) en contacto con diferentes muestras de materiales. Ésta irá soportada en el interior de tubos de acero por los que circulará un gas de purga (He), necesario para arrastrar el tritio generado en el eutéctico y permeado a través de las paredes de las cápsulas (continuamente, durante irradiación). Estos tubos, a su vez, se instalarán en una carcasa también de acero que proporcionará soporte y refrigeración tanto a los tubos como a sus cápsulas experimentales interiores. El módulo, en su conjunto, permitirá la extracción de las señales experimentales y el gas de purga. Así, a través de la estación de medida de tritio y el sistema de control, se obtendrán los datos experimentales para su análisis y extracción de conclusiones experimentales. Además del análisis de datos experimentales, algunas de estas señales tendrán una función de seguridad y por tanto jugarán un papel primordial en la operación del módulo. Para el correcto funcionamiento de las cápsulas y poder controlar su temperatura, cada cápsula se equipará con un calentador eléctrico y por tanto el módulo requerirá también ser conectado a la alimentación eléctrica. El diseño del módulo y su lógica de operación se describe en detalle en esta tesis. La justificación técnica de cada una de las partes que componen el módulo se ha realizado con soporte de cálculos de transporte de tritio, termohidráulicos y mecánicos. Una de las principales conclusiones de los cálculos de transporte de tritio es que es perfectamente viable medir el tritio permeado en las cápsulas mediante cámaras de ionización 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 cápsulas a bajas temperaturas o si llevan recubrimientos antipermeación. Desde un punto de vista de seguridad, el conocimiento de la cantidad de tritio que está siendo transportada con el gas de purga puede ser usado para detectar de ciertos problemas que puedan estar sucediendo en el módulo como por ejemplo, la rotura de una cápsula. Además, es necesario conocer el balance de tritio de la instalación. Las pérdidas esperadas el refrigerante y la celda caliente de IFMIF se pueden considerar despreciables para condiciones normales de funcionamiento. Los cálculos termohidráulicos se han realizado con el objetivo de optimizar el diseño de las cápsulas experimentales y el LBVM de manera que se pueda cumplir el principal requisito del módulo que es llevar a cabo los experimentos a temperaturas comprendidas entre 300-550ºC. Para ello, se ha dimensionado la refrigeración necesaria del módulo y evaluado la geometría de las cápsulas, tubos experimentales y la zona experimental del contenedor. Como consecuencia de los análisis realizados, se han elegido cápsulas y tubos cilíndricos instalados en compartimentos cilíndricos debido a su buen comportamiento mecánico (las tensiones debidas a la presión de los fluidos se ven reducidas significativamente con una geometría cilíndrica en lugar de prismática) y térmico (uniformidad de temperatura en las paredes de los tubos y cápsulas). Se han obtenido campos de presión, temperatura y velocidad en diferentes zonas críticas del módulo concluyendo que la presente propuesta es factible. Cabe destacar que el uso de códigos fluidodinámicos (e.g. ANSYS-CFX, utilizado en esta tesis) para el diseño de cápsulas experimentales de IFMIF no es directo. La razón 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 código han tenido que ser estudiados con detalle y validados con resultados experimentales. El modelo SST (del inglés, “Shear Stress Transport Model”) para turbulencia en transición ha sido identificado como adecuado para simular el comportamiento del helio de refrigeración y la temperatura en las paredes de las cápsulas experimentales. Con la geometría propuesta y los valores principales de refrigeración y purga definidos, se ha analizado el comportamiento mecánico de cada uno de los tubos experimentales que contendrá el módulo. Los resultados de tensiones obtenidos, han sido comparados con los valores máximos recomendados en códigos de diseño estructural como el SDC-IC (del inglés, “Structural Design Criteria for ITER Components”) para así evaluar el grado de protección contra el colapso plástico. La conclusión del estudio muestra que la propuesta es mecánicamente robusta. El LBVM implica el uso de metales líquidos y la generación de tritio además del riesgo asociado a la activación neutrónica. Por ello, se han estudiado los riesgos asociados al uso de metales líquidos y el tritio. Además, se ha incluido una evaluación preliminar de los riesgos radiológicos asociados a la activación de materiales y el calor residual en el módulo después de la irradiación así como un escenario de pérdida de refrigerante. Los riesgos asociados al módulo de naturaleza convencional están asociados al manejo de metales líquidos cuyas reacciones con aire o agua se asocian con emisión de aerosoles y probabilidad de fuego. De entre los riesgos nucleares destacan la generación de gases radiactivos como el tritio u otros radioisótopos volátiles como el Po-210. No se espera que el módulo suponga un impacto medioambiental asociado a posibles escapes. Sin embargo, es necesario un manejo adecuado tanto de las cápsulas experimentales como del módulo contenedor así como de las líneas de purga durante operación. Después de un día de después de la parada, tras un año de irradiación, tendremos una dosis de contacto de 7000 Sv/h en la zona experimental del contenedor, 2300 Sv/h en la cápsula y 25 Sv/h en el LiPb. El uso por lo tanto de manipulación remota está previsto para el manejo del módulo irradiado. Por último, en esta tesis se ha estudiado también las posibilidades existentes para la fabricación del módulo. De entre las técnicas propuestas, destacan la electroerosión, soldaduras por haz de electrones o por soldadura láser. Las bases para el diseño final del LBVM han sido pues establecidas en el marco de este trabajo y han sido incluidas en el diseño intermedio de IFMIF, que será desarrollado en el futuro, como parte del diseño final de la instalación 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 300-550ºC. 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.

Neutron field characteristics of Ciemat's Neutron Standards Laboratory Hector Rene Vega-Carrillo

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.

A strategy for organ allografts without using immunosuppressants or irradiation

A strategy to achieve regular and long lasting organ and tissue allografts without using immunosuppressants and/or irradiation has been established for mice. One hundred percent of skin allografts can be induced to survive >350 days after transplantation if spleen cells from the same donors are first injected into the portal vein of the recipients. The mechanisms underlying this long-term tolerance induction can be described as follows: (i) donor T cells from the spleen of the donor facilitate the acceptance of the allogeneic engraftment, (ii) donor-specific anergy is induced in the cytotoxic T-lymphocytes of the recipients, (iii) T helper type 2 cells become the dominant T cells in the recipients that are accepting the skin transplants, and (iv) a lasting chimerism (microchimerism) is established in these recipients. This strategy, perhaps with minor modifications, might permit one also to overcome major barriers to organ allografting in humans. If this were the case, it could represent production of long lasting immunologic tolerance without need for irradiation or cytotoxic chemo-preparative regimen and as such could greatly facilitate allotransplantation free of episodes of chronic or acute rejection or toxic and damaging preparatory regimens.

Homogeneous immunoconjugates for boron neutron-capture therapy: Design, synthesis, and preliminary characterization

The application of immunoprotein-based targeting strategies to the boron neutron-capture therapy of cancer poses an exceptional challenge, because viable boron neutron-capture therapy by this method will require the efficient delivery of 103 boron-10 atoms by each antigen-binding protein. Our recent investigations in this area have been focused on the development of efficient methods for the assembly of homogeneous immunoprotein conjugates containing the requisite boron load. In this regard, engineered immunoproteins fitted with unique, exposed cysteine residues provide attractive vehicles for site-specific modification. Additionally, homogeneous oligomeric boron-rich phosphodiesters (oligophosphates) have been identified as promising conjugation reagents. The coupling of two such boron-rich oligophosphates to sulfhydryls introduced to the CH2 domain of a chimeric IgG3 has been demonstrated. The resulting boron-rich immunoconjugates are formed efficiently, are readily purified, and have promising in vitro and in vivo characteristics. Encouragingly, these studies showed subtle differences in the properties of the conjugates derived from the two oligophosphate molecules studied, providing a basis for the application of rational design to future work. Such subtle details would not have been as readily discernible in heterogeneous conjugates, thus validating the rigorous experimental design employed here.

Transcription and activity of antioxidant enzymes after ionizing irradiation in radiation-resistant and radiation-sensitive mice

The involvement of the antioxidant enzymes superoxide dismutase (SOD), catalase (CAT), and glutathione peroxidase in radiobiological processes has been described at the enzyme activity level. We irradiated radiation-resistant (RR) and radiation-sensitive (RS) mice and studied antioxidant enzymes at the transcriptional and activity level. In addition, aromatic hydroxylation and lipid peroxidation parameters were determined to study radiation resistance at the oxidation level. RS BALB/c/J Him mice and RR C3H He/Him mice were whole-body-irradiated with x-rays at 2, 4, and 6 Gy and killed 5, 15, and 30 min after irradiation. mRNA was isolated from liver and hybridized with probes for antioxidant enzymes and β-actin as a housekeeping gene control. Antioxidant enzyme activities were determined by standard assays. Parameters for aromatic hydroxylation (o-tyrosine) and lipid peroxidation (malondialdehyde) were determined by HPLC methods. Antioxidant transcription was unchanged in contrast to antioxidant activities; SOD and CAT activities were elevated within 15 min in RR animals but not in RS mice, at all doses studied. Glutathione peroxidase activity was not different between RR and RS mice and was only moderately elevated after irradiation. No significant differences were found between RR and RS animals at the oxidation level, although a radiation dose-dependent increase of oxidation products was detected in both groups. We found that ionizing irradiation led to increased antioxidant activity only minutes after irradiation in the absence of increased transcription of these antioxidant enzymes. RR animals show higher antioxidant enzyme activities than do RS mice, but oxidation products are comparable in RS and RR mice. As unchanged transcription of antioxidant enzymes could not have been responsible for the increased antioxidant enzyme activities, preformed antioxidant enzymes should have been released by the irradiation process. This would be in agreement with previous studies of preformed, stored SOD. The finding of higher SOD and CAT activities in RR than in RS animals could point to a role for these antioxidant enzymes for the process of radiation sensitivity.

The inhibition of antigen-presenting activity of dendritic cells resulting from UV irradiation of murine skin is restored by in vitro photorepair of cyclobutane pyrimidine dimers

Exposing skin to UVB (280–320 nm) radiation suppresses contact hypersensitivity by a mechanism that involves an alteration in the activity of cutaneous antigen-presenting cells (APC). UV-induced DNA damage appears to be an important molecular trigger for this effect. The specific target cells in the skin that sustain DNA damage relevant to the immunosuppressive effect have yet to be identified. We tested the hypothesis that UV-induced DNA damage in the cutaneous APC was responsible for their impaired ability to present antigen after in vivo UV irradiation. Cutaneous APC were collected from the draining lymph nodes of UVB-irradiated, hapten-sensitized mice and incubated in vitro with liposomes containing a photolyase (Photosomes; Applied Genetics, Freeport, NY), which, upon absorption of photoreactivating light, splits UV-induced cyclobutane pyrimidine dimers. Photosome treatment followed by photoreactivating light reduced the number of dimer-containing APC, restored the in vivo antigen-presenting activity of the draining lymph node cells, and blocked the induction of suppressor T cells. Neither Photosomes nor photoreactivating light alone, nor photoreactivating light given before Photosomes, restored APC activity, and Photosome treatment did not reverse the impairment of APC function when isopsoralen plus UVA (320–400 nm) radiation was used instead of UVB. These controls indicate that the restoration of APC function matched the requirements of Photosome-mediated DNA repair for dimers and post-treatment photoreactivating light. These results provide compelling evidence that it is UV-induced DNA damage in cutaneous APC that leads to reduced immune function.

Synthesis and in vivo murine evaluation of Na4[1-(1′-B10H9)-6-SHB10H8] as a potential agent for boron neutron capture therapy

Reaction of the normal isomer of [B20H18]2− and the protected thiol anion, [SC(O)OC(CH3)3]−, produces an unexpected isomer of [B20H17SC(O)OC(CH3)3]4− directly and in good yield. The isomer produced under mild conditions is characterized by an apical–apical boron atom intercage connection as well as the location of the thiol substituent on an equatorial belt adjacent to the terminal boron apex. Although the formation of this isomer from nucleophilic attack of the normal isomer of [B20H18]2− has not been reported previously, the isomeric assignment has been unambiguously confirmed by one-dimensional and two-dimensional 11B NMR spectroscopy. Deprotection of the thiol substituent under acidic conditions produces a protonated intermediate, [B20H18SH]3−, which can be deprotonated with a suitable base to yield the desired product, [B20H17SH]4−. The sodium salt of the resulting [B20H17SH]4− ion has been encapsulated in small, unilamellar liposomes, which are capable of delivering their contents selectively to tumors in vivo, and investigated as a potential agent for boron neutron capture therapy. The biodistribution of boron was determined after intravenous injection of the liposomal suspension into BALB/c mice bearing EMT6 mammary adenocarcinoma. At low injected doses, the tumor boron concentration increased throughout the time-course experiment, resulting in a maximum observed boron concentration of 46.7 μg of B per g of tumor at 48 h and a tumor to blood boron ratio of 7.7. The boron concentration obtained in the tumor corresponds to 22.2% injected dose (i.d.) per g of tissue, a value analogous to the most promising polyhedral borane anions investigated for liposomal delivery and subsequent application in boron neutron capture therapy.