Annealing effects on microstructure and coercive field of ferritic-martensitic ODS Eurofer steel

Oxide dispersion strengthened reduced-activation ferritic-martensitic steels are promising candidates for applications in future fusion power plants. Samples of a reduced activation ferritic-martensitic 9 wt.%Cr-oxide dispersion strengthened Eurofer steel were cold rolled to 80% reduction in thickness and annealed in vacuum for 1 h from 200 to 1350 degrees C to evaluate its thermal stability. Vickers microhardness testing and electron backscatter diffraction (EBSD) were used to characterize the microstructure. The microstructural changes were also followed by magnetic measurements, in particular the corresponding variation of the coercive field (H(c)), as a function of the annealing treatment. Results show that magnetic measurements were sensitive to detect the changes, in particular the martensitic transformation, in samples annealed above 850 degrees C (austenitic regime). (C) 2010 Elsevier B.V. All rights reserved.

Annealing behavior of ferritic-martensitic 9%Cr-ODS-Eurofer steel

Oxide dispersion strengthened ferritic-martensitic steels are potential candidates for applications in future fusion power plants. High creep resistance, good oxidation resistance, reduced neutron activation and microstructural long-term stability at temperatures of about 650-700 degrees C are required in this context. In order to evaluate its thermal stability in the ferritic phase field, samples of the reduced activation ferritic-martensitic 9%Cr-ODS-Eurofer steel were cold rolled to 50% and 80% reductions and further annealed in vacuum from 300 to 800 degrees C for 1 h. The characterization in the annealed state was performed by scanning electron microscopy in the backscattered electron mode, high-resolution electron backscatter diffraction and transmission electron microscopy. Results show that the fine dispersion of Y-based particles (about 10 nm in size) is effective to prevent recrystallization. The low recrystallized volume fraction (<0.1) is associated to the nuclei found at prior grain boundaries and around large M(23)C(6) particles. Static recovery was found to be the predominant softening mechanism of this steel in the investigated temperature range. (c) 2010 Elsevier B.V. All rights reserved.

High-resolution transmission electron microscopy and electron backscatter diffraction in nanoscaled ferritic and ferritic-martensitic oxide dispersion strengthened-steels

For specific blanket and divertor applications in future fusion power reactors a replacement of presently considered reduced activation ferritic martensitic (RAFM) steels as a structural material by suitable oxide dispersion strengthened ferritic martensitic steels would allow a substantial increase of the operating temperature from similar to 823 to about 923 K. Due to this reason the RAFM-alloy ODS-Eurofer has already been developed and produced with industrial partners. In the He-cooled modular divertor concept, where temperatures above 923 K will arise, an ODS-steel with a purely ferritic matrix is advantageous, because of missing phase transitions. Due to this reason, a special ferritic ODS-steel is being manufactured as well. In this work the microstructures of these two ODS-alloy types, analysed mainly by high resolution TEM are compared, with respect to different manufacturing processes. In addition first results of high resolution EBSD scans together with determined orientation maps of the RAFM steel ODS-Eurofer will also be presented. (C) 2008 Elsevier B.V. All rights reserved.

ANNEALING BEHAVIOR OF RAFM ODS-EUROFER STEEL

Oxide-dispersion-strengthened (ODS) ferritic-martensitic steels are candidates for applications in fusion power plants where micro structural long-term stability at temperatures of 650 degrees C to 700 degrees C are required. The microstructural stability of 80% cold-rolled reduced-activation ferritic-martensitic 9% Cr ODS-Eurofer steel was investigated within a wide range of temperatures (300 degrees C to 1350 degrees C). Fine oxide dispersion is very effective to prevent recrystallization in the ferritic phase field. The low recrystallized volume fraction (<0.1) found in samples annealed at 800 degrees C is associated with the nuclei found at prior grain boundaries and around coarse M23C6 particles. The combination of retarding effects such as Zener drag and concurrent recovery decrease the local stored energy and impede further growth of the recrystallization nuclei. Above 90 degrees C, martensitic transformation takes place with consequent coarsening. Significant changes in crystallographic texture are also reported.

Development, mechanical characterization and qualification in liquid lead of stainless steels for structural applications in the future fission and fusion nuclear reactors

The research activity carried out in the Brasimone Research Center of ENEA concerns the development and mechanical characterization of steels conceived as structural materials for future fission reactors (Heavy Liquid Metal IV Generation reactors: MYRRHA and ALFRED) and for the future fusion reactor DEMO. Within this framework, two parallel lines of research have been carried out: (i) characterization in liquid lead of steels and weldings for the components of the IV Generation fission reactors (GIV) by means of creep and SSRT (Slow Strain Rate Tensile) tests; (ii) development and screening on mechanical properties of RAFM (Reduced Activation Ferritic Martensitic) steels to be employed as structural materials of the future DEMO fusion reactor. The doctoral work represents therefore a comprehensive report of the research carried out on nuclear materials both from the point of view of the qualification of existing (commercial) materials for their application in the typical environmental conditions of 4th generation fission reactors operating with lead as coolant, and from the point of view of the metallurgical study (with annexed microstructural and mechanical characterization of the selected compositions / Thermo Mechanical Treatment (TMT) options) of new compositional variants to be proposed for the “Breeding Blanket” of the future DEMO Fusion Reactor.

ANNEALING EFFECTS ON THE MICROSTRUCTURE OF FERRITIC-MARTENSITIC ODS-EUROFER STEEL

Oxide dispersion strengthened (ODS) ferritic/martensitic (FM) steels are promising candidates for structural applications in future fusion power reactors. In order to evaluate the thermal stability of 80% cold-rolled ODS-EUROFER, samples were annealed for 1 h at temperatures up to about 0.9 T(m), where T(m) is the absolute melting point. The characterization of the annealed samples was performed using transmission electron microscopy and electron backscatter diffraction. Results show that static recovery is the main softening mechanism of this steel when annealed below 800 degrees C. The volume fraction of recrystallized grains is quite small (below 0.10). Above 900 degrees C, martensitic transformation takes place causing pronounced hardening. Large M(23)C(6) particles are found at the grain boundaries after tempering at 750 degrees C for 2 h.

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.

Oxidation and tensile behavior of ferritic/martensitic steels after exposure to lead-bismuth eutectic

Two ferritic/martensitic steels, T91 steel and newly developed SIMP steel, were subject to tensile test after being oxidized in the liquid lead-bismuth eutectic (LBE) at 873 K for 500 h, 1000 h and 2000 h. Tensile tests were also carried out on the steels only thermally aged at 873 K. The result shows that thermal aging has no effect. Exposure to LBE at 873 K leads to a slight decrease in strength, but a large decrease in elongation when tested at 873 K. When tested at 873 K after 2000 h exposure, the tensile strength of T91 decreases slightly, and elongation from 39% to 21%. For SIMP, the decreases are slightly and from 44% to 28%, for tensile strength and elongation, respectively. The room temperature strength has slightly larger percentage reductions after the LBE exposure, but the elongation changes little.

Evaluation of austenitic-ferritic stainless steel wires for orthodontic applications

Several studies have shown that austenitic stainless steels are suitable for use in the final phases of orthodontic treatments, such as finishing and retention. These steels demonstrate appropriate mechanical properties, such as high ultimate tensile strength and good corrosion resistance. A new class of materials, the austenic-ferritic stainless steels, is substituting for austenitic stainless steels in several industrial applications where these properties are necessary. This work supports the hypothesis that orthodontic wires of austenic-ferritic stainless steels can replace austenitic stainless steels. The advantages are cost reduction and decrease of the nickel hypersensitivity effect in patients undergoing orthodontic treatments. The object of this study was to evaluate wires of austenitic-ferritic stainless steel SEW 410 Nr. 14517 (Cr26Ni6Mo3Cu3) produced by cold working through rolling and drawing processes. Tests were performed to evaluate the ultimate tensile strength, hardness, ductility, and formability. In accordance with technical standards the wires exhibited ultimate tensile strength and ductility suitable for orthodontic clinical applications. These austenitie-ferritic wires can be an alternative to substitute the common commercial wires of austenic stainless steels with the advantage of decreasing the nickel content.

Cleavage of mitochondrial antiviral signaling protein in the liver of patients with chronic hepatitis C correlates with a reduced activation of the endogenous interferon system.

Hepatitis C virus (HCV) infection induces the endogenous interferon (IFN) system in the liver in some but not all patients with chronic hepatitis C (CHC). Patients with a pre-activated IFN system are less likely to respond to the current standard therapy with pegylated IFN-alpha. Mitochondrial antiviral signaling protein (MAVS) is an important adaptor molecule in a signal transduction pathway that senses viral infections and transcriptionally activates IFN-beta. The HCV NS3-4A protease can cleave and thereby inactivate MAVS in vitro, and, therefore, might be crucial in determining the activation status of the IFN system in the liver of infected patients. We analyzed liver biopsies from 129 patients with CHC to investigate whether MAVS is cleaved in vivo and whether cleavage prevents the induction of the endogenous IFN system. Cleavage of MAVS was detected in 62 of the 129 samples (48%) and was more extensive in patients with a high HCV viral load. MAVS was cleaved by all HCV genotypes (GTs), but more efficiently by GTs 2 and 3 than by GTs 1 and 4. The IFN-induced Janus kinase (Jak)-signal transducer and activator of transcription protein (STAT) pathway was less frequently activated in patients with cleaved MAVS, and there was a significant inverse correlation between cleavage of MAVS and the expression level of the IFN-stimulated genes IFI44L, Viperin, IFI27, USP18, and STAT1. We conclude that the pre-activation status of the endogenous IFN system in the liver of patients with CHC is in part regulated by cleavage of MAVS.

Influence of aluminium on dimensional change of sintered 430 Ferritic Stainless Steel

Aluminium is added to decrease matrix chromium losses on 430 stainless steel sintered on nitrogen atmosphere. Three different ways were used to add a 3% (in weight) aluminium: as elemental powder, as prealloyed powder, and as intermetallic Fe-AI compound. After die pressing at densities between 6.1-6.5 g/cm3, samples were sintered on vacuum and on N2-5%H2 atmosphere in a dilatometric furnace. Therefore, dimensional change was recorded during sintering. Weight gain was obtained after nitrogen sintering on all materials due to nitrides formation. Sample expansion was obtained on all nitrogen sintered steels with Al additions. Microstructure showed a dispersion of aluminium nitrides when pre-alloyed powders are used. On the contrary, aluminium nitride areas can be found when aluminium is added as elemental powders or as Fe-AI intermetallics. Also nitrogen atmosphere leads to austenite formation and hence, on cooling, dilatometric results showed a dimensional change at austenitic-ferritic phase transformation temperature.

Residual stress distribution in ferritic-austenitic steel joints made by laser welding

The present investigation addresse the influence of laser welding process-ing parameters used for joining dis-similar metals (ferritic to austenitic steel), on the induced residual stress field. Welding was performed on a Nd:YAG laser DY033 (3300 W) in a continuous wave (CW), keyhole mode. The base metals (BM) employed in this study are AISI 1010 carbon steel (CS) and AISI 304L austenitic stainless steel (SS). Pairs of dissimilar plates of 200 mm x 45 mm x 3 mm were butt joined by laser welding. Different sets of parameters were used to engineer the base metals apportionment at joint formation, namely distinct dilution rates. Residual strain scanning, carried out by neutron diffraction was used to assess the joints. Through-thickness residual stress maps were determined for the laser welded samples of dis-similar steels using high spatial reso-lution. As a result, an appropriate set of processing parameters, able to mi-nimize the local tensile residual stress associated to the welding process, was found.

Overall vs. local mechanical behaviour of ferritic austenitic steel joints made by laser welding

The present investigation addresses the overall and local mechanical performance of dissimilar joints of low carbon steel (CS) and stainless Steel (SS) thin sheets achieved by laser welding in case of heat source displacement from the weld gap centreline towards CS. Welding was performed on a Nd:YAG laser DY033 (3300 W) in a continuos wave (CW), keyhole mode. The tensile behavior of the joint different zones assessed by using a video-image based system (VIC-2D) reveals that the residual stress field, together with the positive difference in yield between the weld metal and the base materials protects the joint from being plastically deformed. The tensile loadings of flat transverse specimens generate the strain localization and failure in CS, far away from the weld.