Preliminary approach to neutron instrument selecction at ESSB-Bilbao based on experience at ISIS molecular spectroscopy group

Collaborative efforts between the Neutronics and Target Design Group at the Instituto de Fusión Nuclear and the Molecular Spectroscopy Group at the ISIS Pulsed Neutron and Muon Source date back to 2012 in the context of the ESS-Bilbao project. The rationale for these joint activities was twofold, namely: to assess the realm of applicability of the low-energy neutron source proposed by ESS-Bilbao - for details; and to explore instrument capabilities for pulsed-neutron techniques in the range 0.05-3 ms, a time range where ESS-Bilbao and ISIS could offer a significant degree of synergy and complementarity. As part of this collaboration, J.P. de Vicente has spent a three-month period within the ISIS Molecular Spectroscopy Group, to gain hands-on experience on the practical aspects of neutron-instrument design and the requisite neutron-transport simulations. To date, these activities have resulted in a joint MEng thesis as well as a number of publications and contributions to national and international conferences. Building upon these previous works, the primary aim of this report is to provide a self-contained discussion of general criteria for instrument selection at ESS-Bilbao, the first accelerator-driven, low-energy neutron source designed in Spain. To this end, Chapter 1 provides a brief overview of the current design parameters of the accelerator and target station. Neutron moderation is covered in Chapter 2, where we take a closer look at two possible target-moderator-reflector configurations and pay special attention to the spectral and temporal characteristics of the resulting neutron pulses. This discussion provides a necessary starting point to assess the operation of ESSB in short- and long-pulse modes. These considerations are further explored in Chapter 3, dealing with the primary characteristics of ESS-Bilbao as a short- or long-pulse facility in terms of accessible dynamic range and spectral resolution. Other practical aspects including background suppression and the use of fast choppers are also discussed. The guiding principles introduced in the first three chapters are put to use in Chapter 4 where we analyse in some detail the capabilities of a small-angle scattering instrument, as well as how specific scientific requirements can be mapped onto the optimal use of ESS-Bilbao for condensed-matter research. Part 2 of the report contains additional supporting documentation, including a description of the ESSB McStas component, a detailed characterisation of moderator response and neutron pulses, and estimates ofparameters associated with the design and operation of neutron choppers. In closing this brief foreword, we wish to thank both ESS-Bilbao and ISIS for their continuing encouragement and support along the way.

Conceptual design of the beryllium rotating target for the ESS-Bilbao facility

The ESS-Bilbao facility, hosted by the University of the Basque Country (UPV/EHU), envisages the operation of a high-current proton accelerator delivering beams with energies up to 50 MeV. The time-averaged proton current will be 2.25 mA, delivered by 1.5 ms proton pulses with a repetition rate of 20 Hz. This beam will feed a neutron source based upon the Be (p,n) reaction, which will enable the provision of relevant neutron experimentation capabilities. The neutron source baseline concept consists in a rotating beryllium target cooled by water. The target structure will comprise a rotatable disk made of 6061-T6 aluminium alloy holding 20 beryllium plates. Heat dissipation from the target relies upon a distribution of coolant-flow channels. The practical implementation of such a concept is here described with emphasis put on the beryllium plates thermo-mechanical optimization, the chosen coolant distribution system as well as the mechanical behavior of the assembly.

Neutronic design for ESS-Bilbao neutron source

The European Spallation Source-Bilbao (ESS-Bilbao) project plans to build an accelerator facility compliant with the ESS-AB requirements which will be able to drive several experimental stations for research purposes involving intense proton beams with currents up to 75 mA, 50 MeV of final energy, 1.5 ms of pulse length and up to 50 Hz repetition rate. The accelerator will also drive a compact neutron source which will provide useful neutron beams to carry out experiments on moderator optimization, neutron optics devices and general neutron instrumentation as well as preparation work for experiments to be carried out by neutron beam users at the large facilities.

Modelling and Control of Stepper Motors for High Accuracy Positioning Systems Used in Radioactive Environments

Hybrid Stepper Motors are widely used in open-loop position applications. They are the choice of actuation for the collimators in the Large Hadron Collider, the largest particle accelerator at CERN. In this case the positioning requirements and the highly radioactive operating environment are unique. The latter forces both the use of long cables to connect the motors to the drives which act as transmission lines and also prevents the use of standard position sensors. However, reliable and precise operation of the collimators is critical for the machine, requiring the prevention of step loss in the motors and maintenance to be foreseen in case of mechanical degradation. In order to make the above possible, an approach is proposed for the application of an Extended Kalman Filter to a sensorless stepper motor drive, when the motor is separated from its drive by long cables. When the long cables and high frequency pulse width modulated control voltage signals are used together, the electrical signals difer greatly between the motor and drive-side of the cable. Since in the considered case only drive-side data is available, it is therefore necessary to estimate the motor-side signals. Modelling the entire cable and motor system in an Extended Kalman Filter is too computationally intensive for standard embedded real-time platforms. It is, in consequence, proposed to divide the problem into an Extended Kalman Filter, based only on the motor model, and separated motor-side signal estimators, the combination of which is less demanding computationally. The efectiveness of this approach is shown in simulation. Then its validity is experimentally demonstrated via implementation in a DSP based drive. A testbench to test its performance when driving an axis of a Large Hadron Collider collimator is presented along with the results achieved. It is shown that the proposed method is capable of achieving position and load torque estimates which allow step loss to be detected and mechanical degradation to be evaluated without the need for physical sensors. These estimation algorithms often require a precise model of the motor, but the standard electrical model used for hybrid stepper motors is limited when currents, which are high enough to produce saturation of the magnetic circuit, are present. New model extensions are proposed in order to have a more precise model of the motor independently of the current level, whilst maintaining a low computational cost. It is shown that a significant improvement in the model It is achieved with these extensions, and their computational performance is compared to study the cost of model improvement versus computation cost. The applicability of the proposed model extensions is demonstrated via their use in an Extended Kalman Filter running in real-time for closed-loop current control and mechanical state estimation. An additional problem arises from the use of stepper motors. The mechanics of the collimators can wear due to the abrupt motion and torque profiles that are applied by them when used in the standard way, i.e. stepping in open-loop. Closed-loop position control, more specifically Field Oriented Control, would allow smoother profiles, more respectful to the mechanics, to be applied but requires position feedback. As mentioned already, the use of sensors in radioactive environments is very limited for reliability reasons. Sensorless control is a known option but when the speed is very low or zero, as is the case most of the time for the motors used in the LHC collimator, the loss of observability prevents its use. In order to allow the use of position sensors without reducing the long term reliability of the whole system, the possibility to switch from closed to open loop is proposed and validated, allowing the use of closed-loop control when the position sensors function correctly and open-loop when there is a sensor failure. A different approach to deal with the switched drive working with long cables is also presented. Switched mode stepper motor drives tend to have poor performance or even fail completely when the motor is fed through a long cable due to the high oscillations in the drive-side current. The design of a stepper motor output fillter which solves this problem is thus proposed. A two stage filter, one devoted to dealing with the diferential mode and the other with the common mode, is designed and validated experimentally. With this ?lter the drive performance is greatly improved, achieving a positioning repeatability even better than with the drive working without a long cable, the radiated emissions are reduced and the overvoltages at the motor terminals are eliminated.

Evaluation of preisach model of hysteresis to model superconductors

The Large Hadron Collider is the world’s largest and most powerful particle accelerator. The project is divided in phases. The first one goes from 2009 until 2020. The second phase will consist of the implementation of upgrades. One of the upgrades is to increase the ratio of collision, the luminosity. This objective is the main of one of the most important projects which is carrying out the upgrades: Hi-Lumi LHC project. Increasing luminosity could be done by using a new material in the superconductor magnets placed at the interaction points: Nb3Sn, instead of NbTi, the one being used right now. Before implementing it many aspects should be analysed. One of them is the induction magnetic field quality. The tool used so far has been ROXIE, software developed at CERN by S. Russenschuck. One of the main features of the programme is the time-transient analysis, which is based on three mathematical models. It is quite precise for fields above 1.5 Tesla. However, they are not very accurate for lower fields. Therefore the aim of this project is to evaluate a more accurate model: Classical Preisach Model of Hysteresis, in order to better analyse induced field quality in the new material Nb3Sn. Resumen: El Gran Colisionador de Hadrones es el mayor acelerador de partículas circular del mundo. Se trata de uno de los mayores proyectos de investigación. La primera fase de funcionamiento comprende desde 2009 a 2020, cuando comenzará la siguiente fase. Durante el primer periodo se han pensado mejoras para que puedan ser implementadas en la segunda fase. Una de ellas es el aumento del ratio de las colisiones entre protones por choque. Este es el principal objetivo de uno de los proyectos que está llevando a cabo las mejoras a ser implementadas en 2020: Hi- Lumi LHC. Se cambiarán los imanes superconductores de NbTi de las dos zonas principales de interacción, y se sustituirán por imanes de Nb3Sn. Esta sustituciónn conlleva un profundo estudio previo. Entre otros, uno de los factores a analizar es la calidad del campo magnético. La herramienta utilizada es el software desarrollado por S. Russenschuck en el CERN llamado ROXIE. Está basado en tres modelos de magnetización, los cuales son precisos para campos mayores de 1.5 T. Sin embargo, no lo son tanto para campos menores. Con este proyecto se pretende evaluar la implementación de un cuarto modelo, el modelo clásico de histéresis de Preisach que permita llevar a cabo un mejor análisis de la calidad del campo inducido por el futuro material a utilizar en algunos de los imanes.

Monte Carlo Neutronics and Thermal-hydraulics coupling applied to Fast Reactors

Un escenario habitualmente considerado para el uso sostenible y prolongado de la energía nuclear contempla un parque de reactores rápidos refrigerados por metales líquidos (LMFR) dedicados al reciclado de Pu y la transmutación de actínidos minoritarios (MA). Otra opción es combinar dichos reactores con algunos sistemas subcríticos asistidos por acelerador (ADS), exclusivamente destinados a la eliminación de MA. El diseño y licenciamiento de estos reactores innovadores requiere herramientas computacionales prácticas y precisas, que incorporen el conocimiento obtenido en la investigación experimental de nuevas configuraciones de reactores, materiales y sistemas. A pesar de que se han construido y operado un cierto número de reactores rápidos a nivel mundial, la experiencia operacional es todavía reducida y no todos los transitorios se han podido entender completamente. Por tanto, los análisis de seguridad de nuevos LMFR están basados fundamentalmente en métodos deterministas, al contrario que las aproximaciones modernas para reactores de agua ligera (LWR), que se benefician también de los métodos probabilistas. La aproximación más usada en los estudios de seguridad de LMFR es utilizar una variedad de códigos, desarrollados a base de distintas teorías, en busca de soluciones integrales para los transitorios e incluyendo incertidumbres. En este marco, los nuevos códigos para cálculos de mejor estimación ("best estimate") que no incluyen aproximaciones conservadoras, son de una importancia primordial para analizar estacionarios y transitorios en reactores rápidos. Esta tesis se centra en el desarrollo de un código acoplado para realizar análisis realistas en reactores rápidos críticos aplicando el método de Monte Carlo. Hoy en día, dado el mayor potencial de recursos computacionales, los códigos de transporte neutrónico por Monte Carlo se pueden usar de manera práctica para realizar cálculos detallados de núcleos completos, incluso de elevada heterogeneidad material. Además, los códigos de Monte Carlo se toman normalmente como referencia para los códigos deterministas de difusión en multigrupos en aplicaciones con reactores rápidos, porque usan secciones eficaces punto a punto, un modelo geométrico exacto y tienen en cuenta intrínsecamente la dependencia angular de flujo. En esta tesis se presenta una metodología de acoplamiento entre el conocido código MCNP, que calcula la generación de potencia en el reactor, y el código de termohidráulica de subcanal COBRA-IV, que obtiene las distribuciones de temperatura y densidad en el sistema. COBRA-IV es un código apropiado para aplicaciones en reactores rápidos ya que ha sido validado con resultados experimentales en haces de barras con sodio, incluyendo las correlaciones más apropiadas para metales líquidos. En una primera fase de la tesis, ambos códigos se han acoplado en estado estacionario utilizando un método iterativo con intercambio de archivos externos. El principal problema en el acoplamiento neutrónico y termohidráulico en estacionario con códigos de Monte Carlo es la manipulación de las secciones eficaces para tener en cuenta el ensanchamiento Doppler cuando la temperatura del combustible aumenta. Entre todas las opciones disponibles, en esta tesis se ha escogido la aproximación de pseudo materiales, y se ha comprobado que proporciona resultados aceptables en su aplicación con reactores rápidos. Por otro lado, los cambios geométricos originados por grandes gradientes de temperatura en el núcleo de reactores rápidos resultan importantes para la neutrónica como consecuencia del elevado recorrido libre medio del neutrón en estos sistemas. Por tanto, se ha desarrollado un módulo adicional que simula la geometría del reactor en caliente y permite estimar la reactividad debido a la expansión del núcleo en un transitorio. éste módulo calcula automáticamente la longitud del combustible, el radio de la vaina, la separación de los elementos de combustible y el radio de la placa soporte en función de la temperatura. éste efecto es muy relevante en transitorios sin inserción de bancos de parada. También relacionado con los cambios geométricos, se ha implementado una herramienta que, automatiza el movimiento de las barras de control en busca d la criticidad del reactor, o bien calcula el valor de inserción axial las barras de control. Una segunda fase en la plataforma de cálculo que se ha desarrollado es la simulació dinámica. Puesto que MCNP sólo realiza cálculos estacionarios para sistemas críticos o supercríticos, la solución más directa que se propone sin modificar el código fuente de MCNP es usar la aproximación de factorización de flujo, que resuelve por separado la forma del flujo y la amplitud. En este caso se han estudiado en profundidad dos aproximaciones: adiabática y quasiestática. El método adiabático usa un esquema de acoplamiento que alterna en el tiempo los cálculos neutrónicos y termohidráulicos. MCNP calcula el modo fundamental de la distribución de neutrones y la reactividad al final de cada paso de tiempo, y COBRA-IV calcula las propiedades térmicas en el punto intermedio de los pasos de tiempo. La evolución de la amplitud de flujo se calcula resolviendo las ecuaciones de cinética puntual. Este método calcula la reactividad estática en cada paso de tiempo que, en general, difiere de la reactividad dinámica que se obtendría con la distribución de flujo exacta y dependiente de tiempo. No obstante, para entornos no excesivamente alejados de la criticidad ambas reactividades son similares y el método conduce a resultados prácticos aceptables. Siguiendo esta línea, se ha desarrollado después un método mejorado para intentar tener en cuenta el efecto de la fuente de neutrones retardados en la evolución de la forma del flujo durante el transitorio. El esquema consiste en realizar un cálculo cuasiestacionario por cada paso de tiempo con MCNP. La simulación cuasiestacionaria se basa EN la aproximación de fuente constante de neutrones retardados, y consiste en dar un determinado peso o importancia a cada ciclo computacial del cálculo de criticidad con MCNP para la estimación del flujo final. Ambos métodos se han verificado tomando como referencia los resultados del código de difusión COBAYA3 frente a un ejercicio común y suficientemente significativo. Finalmente, con objeto de demostrar la posibilidad de uso práctico del código, se ha simulado un transitorio en el concepto de reactor crítico en fase de diseño MYRRHA/FASTEF, de 100 MW de potencia térmica y refrigerado por plomo-bismuto. ABSTRACT Long term sustainable nuclear energy scenarios envisage a fleet of Liquid Metal Fast Reactors (LMFR) for the Pu recycling and minor actinides (MAs) transmutation or combined with some accelerator driven systems (ADS) just for MAs elimination. Design and licensing of these innovative reactor concepts require accurate computational tools, implementing the knowledge obtained in experimental research for new reactor configurations, materials and associated systems. Although a number of fast reactor systems have already been built, the operational experience is still reduced, especially for lead reactors, and not all the transients are fully understood. The safety analysis approach for LMFR is therefore based only on deterministic methods, different from modern approach for Light Water Reactors (LWR) which also benefit from probabilistic methods. Usually, the approach adopted in LMFR safety assessments is to employ a variety of codes, somewhat different for the each other, to analyze transients looking for a comprehensive solution and including uncertainties. In this frame, new best estimate simulation codes are of prime importance in order to analyze fast reactors steady state and transients. This thesis is focused on the development of a coupled code system for best estimate analysis in fast critical reactor. Currently due to the increase in the computational resources, Monte Carlo methods for neutrons transport can be used for detailed full core calculations. Furthermore, Monte Carlo codes are usually taken as reference for deterministic diffusion multigroups codes in fast reactors applications because they employ point-wise cross sections in an exact geometry model and intrinsically account for directional dependence of the ux. The coupling methodology presented here uses MCNP to calculate the power deposition within the reactor. The subchannel code COBRA-IV calculates the temperature and density distribution within the reactor. COBRA-IV is suitable for fast reactors applications because it has been validated against experimental results in sodium rod bundles. The proper correlations for liquid metal applications have been added to the thermal-hydraulics program. Both codes are coupled at steady state using an iterative method and external files exchange. The main issue in the Monte Carlo/thermal-hydraulics steady state coupling is the cross section handling to take into account Doppler broadening when temperature rises. Among every available options, the pseudo materials approach has been chosen in this thesis. This approach obtains reasonable results in fast reactor applications. Furthermore, geometrical changes caused by large temperature gradients in the core, are of major importance in fast reactor due to the large neutron mean free path. An additional module has therefore been included in order to simulate the reactor geometry in hot state or to estimate the reactivity due to core expansion in a transient. The module automatically calculates the fuel length, cladding radius, fuel assembly pitch and diagrid radius with the temperature. This effect will be crucial in some unprotected transients. Also related to geometrical changes, an automatic control rod movement feature has been implemented in order to achieve a just critical reactor or to calculate control rod worth. A step forward in the coupling platform is the dynamic simulation. Since MCNP performs only steady state calculations for critical systems, the more straight forward option without modifying MCNP source code, is to use the flux factorization approach solving separately the flux shape and amplitude. In this thesis two options have been studied to tackle time dependent neutronic simulations using a Monte Carlo code: adiabatic and quasistatic methods. The adiabatic methods uses a staggered time coupling scheme for the time advance of neutronics and the thermal-hydraulics calculations. MCNP computes the fundamental mode of the neutron flux distribution and the reactivity at the end of each time step and COBRA-IV the thermal properties at half of the the time steps. To calculate the flux amplitude evolution a solver of the point kinetics equations is used. This method calculates the static reactivity in each time step that in general is different from the dynamic reactivity calculated with the exact flux distribution. Nevertheless, for close to critical situations, both reactivities are similar and the method leads to acceptable practical results. In this line, an improved method as an attempt to take into account the effect of delayed neutron source in the transient flux shape evolutions is developed. The scheme performs a quasistationary calculation per time step with MCNP. This quasistationary simulations is based con the constant delayed source approach, taking into account the importance of each criticality cycle in the final flux estimation. Both adiabatic and quasistatic methods have been verified against the diffusion code COBAYA3, using a theoretical kinetic exercise. Finally, a transient in a critical 100 MWth lead-bismuth-eutectic reactor concept is analyzed using the adiabatic method as an application example in a real system.

Task-based telemanipulation for maintenance in large scientific facilities

This thesis presents a task-oriented approach to telemanipulation for maintenance in large scientific facilities, with specific focus on the particle accelerator facilities at European Organization for Nuclear Research (CERN) in Geneva, Switzerland and GSI Helmholtz Centre for Heavy Ion Research (GSI) in Darmstadt, Germany. It examines how telemanipulation can be used in these facilities and reviews how this differs from the representation of telemanipulation tasks within the literature. It provides methods to assess and compare telemanipulation procedures as well a test suite to compare telemanipulators themselves from a dexterity perspective. It presents a formalisation of telemanipulation procedures into a hierarchical model which can be then used as a basis to aid maintenance engineers in assessing tasks for telemanipulation, and as the basis for future research. The model introduces a new concept of Elemental Actions as the building block of telemanipulation movements and incorporates the dependent factors for procedures at a higher level of abstraction. In order to gain insight into realistic tasks performed by telemanipulation systems within both industrial and research environments a survey of teleoperation experts is presented. Analysis of the responses is performed from which it is concluded that there is a need within the robotics community for physical benchmarking tests which are geared towards evaluating the dexterity of telemanipulators for comparison of their dexterous abilities. A three stage test suite is presented which is designed to allow maintenance engineers to assess different telemanipulators for their dexterity. This incorporates general characteristics of the system, a method to compare kinematic reachability of multiple telemanipulators and physical test setups to assess dexterity from a both a qualitative perspective and measurably by using performance metrics. Finally, experimental results are provided for the application of the proposed test suite onto two telemanipulation systems, one from a research setting and the other within CERN. It describes the procedure performed and discusses comparisons between the two systems, as well as providing input from the expert operator of the CERN system.

Wireless communication enhancement methods for mobile robots in radiation environments

En entornos hostiles tales como aquellas instalaciones científicas donde la radiación ionizante es el principal peligro, el hecho de reducir las intervenciones humanas mediante el incremento de las operaciones robotizadas está siendo cada vez más de especial interés. CERN, la Organización Europea para la Investigación Nuclear, tiene alrededor de unos 50 km de superficie subterránea donde robots móviles controlador de forma remota podrían ayudar en su funcionamiento, por ejemplo, a la hora de llevar a cabo inspecciones remotas sobre radiación en los diferentes áreas destinados al efecto. No solo es preciso considerar que los robots deben ser capaces de recorrer largas distancias y operar durante largos periodos de tiempo, sino que deben saber desenvolverse en los correspondientes túneles subterráneos, tener en cuenta la presencia de campos electromagnéticos, radiación ionizante, etc. y finalmente, el hecho de que los robots no deben interrumpir el funcionamiento de los aceleradores. El hecho de disponer de un sistema de comunicaciones inalámbrico fiable y robusto es esencial para la correcta ejecución de las misiones que los robots deben afrontar y por supuesto, para evitar tales situaciones en las que es necesario la recuperación manual de los robots al agotarse su energía o al perder el enlace de comunicaciones. El objetivo de esta Tesis es proveer de las directrices y los medios necesarios para reducir el riesgo de fallo en la misión y maximizar las capacidades de los robots móviles inalámbricos los cuales disponen de almacenamiento finito de energía al trabajar en entornos peligrosos donde no se dispone de línea de vista directa. Para ello se proponen y muestran diferentes estrategias y métodos de comunicación inalámbrica. Teniendo esto en cuenta, se presentan a continuación los objetivos de investigación a seguir a lo largo de la Tesis: predecir la cobertura de comunicaciones antes y durante las misiones robotizadas; optimizar la capacidad de red inalámbrica de los robots móviles con respecto a su posición; y mejorar el rango operacional de esta clase de robots. Por su parte, las contribuciones a la Tesis se citan más abajo. El primer conjunto de contribuciones son métodos novedosos para predecir el consumo de energía y la autonomía en la comunicación antes y después de disponer de los robots en el entorno seleccionado. Esto es importante para proporcionar conciencia de la situación del robot y evitar fallos en la misión. El consumo de energía se predice usando una estrategia propuesta la cual usa modelos de consumo provenientes de diferentes componentes en un robot. La predicción para la cobertura de comunicaciones se desarrolla usando un nuevo filtro de RSS (Radio Signal Strength) y técnicas de estimación con la ayuda de Filtros de Kalman. El segundo conjunto de contribuciones son métodos para optimizar el rango de comunicaciones usando novedosas técnicas basadas en muestreo espacial que son robustas frente a ruidos de campos de detección y radio y que proporcionan redundancia. Se emplean métodos de diferencia central finitos para determinar los gradientes 2D RSS y se usa la movilidad del robot para optimizar el rango de comunicaciones y la capacidad de red. Este método también se valida con un caso de estudio centrado en la teleoperación háptica de robots móviles inalámbricos. La tercera contribución es un algoritmo robusto y estocástico descentralizado para la optimización de la posición al considerar múltiples robots autónomos usados principalmente para extender el rango de comunicaciones desde la estación de control al robot que está desarrollando la tarea. Todos los métodos y algoritmos propuestos se verifican y validan usando simulaciones y experimentos de campo con variedad de robots móviles disponibles en CERN. En resumen, esta Tesis ofrece métodos novedosos y demuestra su uso para: predecir RSS; optimizar la posición del robot; extender el rango de las comunicaciones inalámbricas; y mejorar las capacidades de red de los robots móviles inalámbricos para su uso en aplicaciones dentro de entornos peligrosos, que como ya se mencionó anteriormente, se destacan las instalaciones científicas con emisión de radiación ionizante. En otros términos, se ha desarrollado un conjunto de herramientas para mejorar, facilitar y hacer más seguras las misiones de los robots en entornos hostiles. Esta Tesis demuestra tanto en teoría como en práctica que los robots móviles pueden mejorar la calidad de las comunicaciones inalámbricas mediante la profundización en el estudio de su movilidad para optimizar dinámicamente sus posiciones y mantener conectividad incluso cuando no existe línea de vista. Los métodos desarrollados en la Tesis son especialmente adecuados para su fácil integración en robots móviles y pueden ser aplicados directamente en la capa de aplicación de la red inalámbrica. ABSTRACT In hostile environments such as in scientific facilities where ionising radiation is a dominant hazard, reducing human interventions by increasing robotic operations are desirable. CERN, the European Organization for Nuclear Research, has around 50 km of underground scientific facilities, where wireless mobile robots could help in the operation of the accelerator complex, e.g. in conducting remote inspections and radiation surveys in different areas. The main challenges to be considered here are not only that the robots should be able to go over long distances and operate for relatively long periods, but also the underground tunnel environment, the possible presence of electromagnetic fields, radiation effects, and the fact that the robots shall in no way interrupt the operation of the accelerators. Having a reliable and robust wireless communication system is essential for successful execution of such robotic missions and to avoid situations of manual recovery of the robots in the event that the robot runs out of energy or when the robot loses its communication link. The goal of this thesis is to provide means to reduce risk of mission failure and maximise mission capabilities of wireless mobile robots with finite energy storage capacity working in a radiation environment with non-line-of-sight (NLOS) communications by employing enhanced wireless communication methods. Towards this goal, the following research objectives are addressed in this thesis: predict the communication range before and during robotic missions; optimise and enhance wireless communication qualities of mobile robots by using robot mobility and employing multi-robot network. This thesis provides introductory information on the infrastructures where mobile robots will need to operate, the tasks to be carried out by mobile robots and the problems encountered in these environments. The reporting of research work carried out to improve wireless communication comprises an introduction to the relevant radio signal propagation theory and technology followed by explanation of the research in the following stages: An analysis of the wireless communication requirements for mobile robot for different tasks in a selection of CERN facilities; predictions of energy and communication autonomies (in terms of distance and time) to reduce risk of energy and communication related failures during missions; autonomous navigation of a mobile robot to find zone(s) of maximum radio signal strength to improve communication coverage area; and autonomous navigation of one or more mobile robots acting as mobile wireless relay (repeater) points in order to provide a tethered wireless connection to a teleoperated mobile robot carrying out inspection or radiation monitoring activities in a challenging radio environment. The specific contributions of this thesis are outlined below. The first sets of contributions are novel methods for predicting the energy autonomy and communication range(s) before and after deployment of the mobile robots in the intended environments. This is important in order to provide situational awareness and avoid mission failures. The energy consumption is predicted by using power consumption models of different components in a mobile robot. This energy prediction model will pave the way for choosing energy-efficient wireless communication strategies. The communication range prediction is performed using radio signal propagation models and applies radio signal strength (RSS) filtering and estimation techniques with the help of Kalman filters and Gaussian process models. The second set of contributions are methods to optimise the wireless communication qualities by using novel spatial sampling based techniques that are robust to sensing and radio field noises and provide redundancy features. Central finite difference (CFD) methods are employed to determine the 2-D RSS gradients and use robot mobility to optimise the communication quality and the network throughput. This method is also validated with a case study application involving superior haptic teleoperation of wireless mobile robots where an operator from a remote location can smoothly navigate a mobile robot in an environment with low-wireless signals. The third contribution is a robust stochastic position optimisation algorithm for multiple autonomous relay robots which are used for wireless tethering of radio signals and thereby to enhance the wireless communication qualities. All the proposed methods and algorithms are verified and validated using simulations and field experiments with a variety of mobile robots available at CERN. In summary, this thesis offers novel methods and demonstrates their use to predict energy autonomy and wireless communication range, optimise robots position to improve communication quality and enhance communication range and wireless network qualities of mobile robots for use in applications in hostile environmental characteristics such as scientific facilities emitting ionising radiations. In simpler terms, a set of tools are developed in this thesis for improving, easing and making safer robotic missions in hostile environments. This thesis validates both in theory and experiments that mobile robots can improve wireless communication quality by exploiting robots mobility to dynamically optimise their positions and maintain connectivity even when the (radio signal) environment possess non-line-of-sight characteristics. The methods developed in this thesis are well-suited for easier integration in mobile robots and can be applied directly at the application layer of the wireless network. The results of the proposed methods have outperformed other comparable state-of-the-art methods.

Analysis of advanced European nuclear fuel cycle scenarios including transmutation and economic estimates

Four European fuel cycle scenarios involving transmutation options (in coherence with PATEROS and CPESFR EU projects) have been addressed from a point of view of resources utilization and economic estimates. Scenarios include: (i) the current fleet using Light Water Reactor (LWR) technology and open fuel cycle, (ii) full replacement of the initial fleet with Fast Reactors (FR) burning U?Pu MOX fuel, (iii) closed fuel cycle with Minor Actinide (MA) transmutation in a fraction of the FR fleet, and (iv) closed fuel cycle with MA transmutation in dedicated Accelerator Driven Systems (ADS). All scenarios consider an intermediate period of GEN-III+ LWR deployment and they extend for 200 years, looking for long term equilibrium mass flow achievement. The simulations were made using the TR_EVOL code, capable to assess the management of the nuclear mass streams in the scenario as well as economics for the estimation of the levelized cost of electricity (LCOE) and other costs. Results reveal that all scenarios are feasible according to nuclear resources demand (natural and depleted U, and Pu). Additionally, we have found as expected that the FR scenario reduces considerably the Pu inventory in repositories compared to the reference scenario. The elimination of the LWR MA legacy requires a maximum of 55% fraction (i.e., a peak value of 44 FR units) of the FR fleet dedicated to transmutation (MA in MOX fuel, homogeneous transmutation) or an average of 28 units of ADS plants (i.e., a peak value of 51 ADS units). Regarding the economic analysis, the main usefulness of the provided economic results is for relative comparison of scenarios and breakdown of LCOE contributors rather than provision of absolute values, as technological readiness levels are low for most of the advanced fuel cycle stages. The obtained estimations show an increase of LCOE ? averaged over the whole period ? with respect to the reference open cycle scenario of 20% for Pu management scenario and around 35% for both transmutation scenarios. The main contribution to LCOE is the capital costs of new facilities, quantified between 60% and 69% depending on the scenario. An uncertainty analysis is provided around assumed low and high values of processes and technologies.

Sistema de simulación de escenarios de ciclos avanzados de combustible nuclear en transición = Simulation system for advanced fuel cycle transition scenarios

El estudio de los ciclos del combustible nuclear requieren de herramientas computacionales o "códigos" versátiles para dar respuestas al problema multicriterio de evaluar los actuales ciclos o las capacidades de las diferentes estrategias y escenarios con potencial de desarrollo en a nivel nacional, regional o mundial. Por otra parte, la introducción de nuevas tecnologías para reactores y procesos industriales hace que los códigos existentes requieran nuevas capacidades para evaluar la transición del estado actual del ciclo del combustible hacia otros más avanzados y sostenibles. Brevemente, esta tesis se centra en dar respuesta a las principales preguntas, en términos económicos y de recursos, al análisis de escenarios de ciclos de combustible, en particular, para el análisis de los diferentes escenarios del ciclo del combustible de relativa importancia para España y Europa. Para alcanzar este objetivo ha sido necesaria la actualización y el desarrollo de nuevas capacidades del código TR_EVOL (Transition Evolution code). Este trabajo ha sido desarrollado en el Programa de Innovación Nuclear del CIEMAT desde el año 2010. Esta tesis se divide en 6 capítulos. El primer capítulo ofrece una visión general del ciclo de combustible nuclear, sus principales etapas y los diferentes tipos utilizados en la actualidad o en desarrollo para el futuro. Además, se describen las fuentes de material nuclear que podrían ser utilizadas como combustible (uranio y otros). También se puntualizan brevemente una serie de herramientas desarrolladas para el estudio de estos ciclos de combustible nuclear. El capítulo 2 está dirigido a dar una idea básica acerca de los costes involucrados en la generación de electricidad mediante energía nuclear. Aquí se presentan una clasificación de estos costos y sus estimaciones, obtenidas en la bibliografía, y que han sido evaluadas y utilizadas en esta tesis. Se ha incluido también una breve descripción del principal indicador económico utilizado en esta tesis, el “coste nivelado de la electricidad”. El capítulo 3 se centra en la descripción del código de simulación desarrollado para el estudio del ciclo del combustible nuclear, TR_EVOL, que ha sido diseñado para evaluar diferentes opciones de ciclos de combustibles. En particular, pueden ser evaluados las diversos reactores con, posiblemente, diferentes tipos de combustibles y sus instalaciones del ciclo asociadas. El módulo de evaluaciones económica de TR_EVOL ofrece el coste nivelado de la electricidad haciendo uso de las cuatro fuentes principales de información económica y de la salida del balance de masas obtenido de la simulación del ciclo en TR_EVOL. Por otra parte, la estimación de las incertidumbres en los costes también puede ser efectuada por el código. Se ha efectuado un proceso de comprobación cruzada de las funcionalidades del código y se descrine en el Capítulo 4. El proceso se ha aplicado en cuatro etapas de acuerdo con las características más importantes de TR_EVOL, balance de masas, composición isotópica y análisis económico. Así, la primera etapa ha consistido en el balance masas del ciclo de combustible nuclear actual de España. La segunda etapa se ha centrado en la comprobación de la composición isotópica del flujo de masas mediante el la simulación del ciclo del combustible definido en el proyecto CP-ESFR UE. Las dos últimas etapas han tenido como objetivo validar el módulo económico. De este modo, en la tercera etapa han sido evaluados los tres principales costes (financieros, operación y mantenimiento y de combustible) y comparados con los obtenidos por el proyecto ARCAS, omitiendo los costes del fin del ciclo o Back-end, los que han sido evaluado solo en la cuarta etapa, haciendo uso de costes unitarios y parámetros obtenidos a partir de la bibliografía. En el capítulo 5 se analizan dos grupos de opciones del ciclo del combustible nuclear de relevante importancia, en términos económicos y de recursos, para España y Europa. Para el caso español, se han simulado dos grupos de escenarios del ciclo del combustible, incluyendo estrategias de reproceso y extensión de vida de los reactores. Este análisis se ha centrado en explorar las ventajas y desventajas de reprocesado de combustible irradiado en un país con una “relativa” pequeña cantidad de reactores nucleares. Para el grupo de Europa se han tratado cuatro escenarios, incluyendo opciones de transmutación. Los escenarios incluyen los reactores actuales utilizando la tecnología reactor de agua ligera y ciclo abierto, un reemplazo total de los reactores actuales con reactores rápidos que queman combustible U-Pu MOX y dos escenarios del ciclo del combustible con transmutación de actínidos minoritarios en una parte de los reactores rápidos o en sistemas impulsados por aceleradores dedicados a transmutación. Finalmente, el capítulo 6 da las principales conclusiones obtenidas de esta tesis y los trabajos futuros previstos en el campo del análisis de ciclos de combustible nuclear. ABSTRACT The study of the nuclear fuel cycle requires versatile computational tools or “codes” to provide answers to the multicriteria problem of assessing current nuclear fuel cycles or the capabilities of different strategies and scenarios with potential development in a country, region or at world level. Moreover, the introduction of new technologies for reactors and industrial processes makes the existing codes to require new capabilities to assess the transition from current status of the fuel cycle to the more advanced and sustainable ones. Briefly, this thesis is focused in providing answers to the main questions about resources and economics in fuel cycle scenario analyses, in particular for the analysis of different fuel cycle scenarios with relative importance for Spain and Europe. The upgrade and development of new capabilities of the TR_EVOL code (Transition Evolution code) has been necessary to achieve this goal. This work has been developed in the Nuclear Innovation Program at CIEMAT since year 2010. This thesis is divided in 6 chapters. The first one gives an overview of the nuclear fuel cycle, its main stages and types currently used or in development for the future. Besides the sources of nuclear material that could be used as fuel (uranium and others) are also viewed here. A number of tools developed for the study of these nuclear fuel cycles are also briefly described in this chapter. Chapter 2 is aimed to give a basic idea about the cost involved in the electricity generation by means of the nuclear energy. The main classification of these costs and their estimations given by bibliography, which have been evaluated in this thesis, are presented. A brief description of the Levelized Cost of Electricity, the principal economic indicator used in this thesis, has been also included. Chapter 3 is focused on the description of the simulation tool TR_EVOL developed for the study of the nuclear fuel cycle. TR_EVOL has been designed to evaluate different options for the fuel cycle scenario. In particular, diverse nuclear power plants, having possibly different types of fuels and the associated fuel cycle facilities can be assessed. The TR_EVOL module for economic assessments provides the Levelized Cost of Electricity making use of the TR_EVOL mass balance output and four main sources of economic information. Furthermore, uncertainties assessment can be also carried out by the code. A cross checking process of the performance of the code has been accomplished and it is shown in Chapter 4. The process has been applied in four stages according to the most important features of TR_EVOL. Thus, the first stage has involved the mass balance of the current Spanish nuclear fuel cycle. The second stage has been focused in the isotopic composition of the mass flow using the fuel cycle defined in the EU project CP-ESFR. The last two stages have been aimed to validate the economic module. In the third stage, the main three generation costs (financial cost, O&M and fuel cost) have been assessed and compared to those obtained by ARCAS project, omitting the back-end costs. This last cost has been evaluated alone in the fourth stage, making use of some unit cost and parameters obtained from the bibliography. In Chapter 5 two groups of nuclear fuel cycle options with relevant importance for Spain and Europe are analyzed in economic and resources terms. For the Spanish case, two groups of fuel cycle scenarios have been simulated including reprocessing strategies and life extension of the current reactor fleet. This analysis has been focused on exploring the advantages and disadvantages of spent fuel reprocessing in a country with relatively small amount of nuclear power plants. For the European group, four fuel cycle scenarios involving transmutation options have been addressed. Scenarios include the current fleet using Light Water Reactor technology and open fuel cycle, a full replacement of the initial fleet with Fast Reactors burning U-Pu MOX fuel and two fuel cycle scenarios with Minor Actinide transmutation in a fraction of the FR fleet or in dedicated Accelerator Driven Systems. Finally, Chapter 6 gives the main conclusions obtained from this thesis and the future work foreseen in the field of nuclear fuel cycle analysis.

Evaluation of reactivity monitoring techniques at the Yalina – Booster subcritical facility

La gestión de los residuos radiactivos de vida larga producidos en los reactores nucleares constituye uno de los principales desafíos de la tecnología nuclear en la actualidad. Una posible opción para su gestión es la transmutación de los nucleidos de vida larga en otros de vida más corta. Los sistemas subcríticos guiados por acelerador (ADS por sus siglas en inglés) son una de las tecnologías en desarrollo para logar este objetivo. Un ADS consiste en un reactor nuclear subcrítico mantenido en un estado estacionario mediante una fuente externa de neutrones guiada por un acelerador de partículas. El interés de estos sistemas radica en su capacidad para ser cargados con combustibles que tengan contenidos de actínidos minoritarios mayores que los reactores críticos convencionales, y de esta manera, incrementar las tasas de trasmutación de estos elementos, que son los principales responsables de la radiotoxicidad a largo plazo de los residuos nucleares. Uno de los puntos clave que han sido identificados para la operación de un ADS a escala industrial es la necesidad de monitorizar continuamente la reactividad del sistema subcrítico durante la operación. Por esta razón, desde los años 1990 se han realizado varios experimentos en conjuntos subcríticos de potencia cero (MUSE, RACE, KUCA, Yalina, GUINEVERE/FREYA) con el fin de validar experimentalmente estas técnicas. En este contexto, la presente tesis se ocupa de la validación de técnicas de monitorización de la reactividad en el conjunto subcrítico Yalina-Booster. Este conjunto pertenece al Joint Institute for Power and Nuclear Research (JIPNR-Sosny) de la Academia Nacional de Ciencias de Bielorrusia. Dentro del proyecto EUROTRANS del 6º Programa Marco de la UE, en el año 2008 se ha realizado una serie de experimentos en esta instalación concernientes a la monitorización de la reactividad bajo la dirección del CIEMAT. Se han realizado dos tipos de experimentos: experimentos con una fuente de neutrones pulsada (PNS) y experimentos con una fuente continua con interrupciones cortas (beam trips). En el caso de los primeros, experimentos con fuente pulsada, existen dos técnicas fundamentales para medir la reactividad, conocidas como la técnica del ratio bajo las áreas de los neutrones inmediatos y retardados (o técnica de Sjöstrand) y la técnica de la constante de decaimiento de los neutrones inmediatos. Sin embargo, varios experimentos han mostrado la necesidad de aplicar técnicas de corrección para tener en cuenta los efectos espaciales y energéticos presentes en un sistema real y obtener valores precisos de la reactividad. En esta tesis, se han investigado estas correcciones mediante simulaciones del sistema con el código de Montecarlo MCNPX. Esta investigación ha servido también para proponer una versión generalizada de estas técnicas donde se buscan relaciones entre la reactividad el sistema y las cantidades medidas a través de simulaciones de Monte Carlo. El segundo tipo de experimentos, experimentos con una fuente continua e interrupciones del haz, es más probable que sea empleado en un ADS industrial. La versión generalizada de las técnicas desarrolladas para los experimentos con fuente pulsada también ha sido aplicada a los resultados de estos experimentos. Además, el trabajo presentado en esta tesis es la primera vez, en mi conocimiento, en que la reactividad de un sistema subcrítico se monitoriza durante la operación con tres técnicas simultáneas: la técnica de la relación entre la corriente y el flujo (current-to-flux), la técnica de desconexión rápida de la fuente (source-jerk) y la técnica del decaimiento de los neutrones inmediatos. Los casos analizados incluyen la variación rápida de la reactividad del sistema (inserción y extracción de las barras de control) y la variación rápida de la fuente de neutrones (interrupción larga del haz y posterior recuperación). ABSTRACT The management of long-lived radioactive wastes produced by nuclear reactors constitutes one of the main challenges of nuclear technology nowadays. A possible option for its management consists in the transmutation of long lived nuclides into shorter lived ones. Accelerator Driven Subcritical Systems (ADS) are one of the technologies in development to achieve this goal. An ADS consists in a subcritical nuclear reactor maintained in a steady state by an external neutron source driven by a particle accelerator. The interest of these systems lays on its capacity to be loaded with fuels having larger contents of minor actinides than conventional critical reactors, and in this way, increasing the transmutation rates of these elements, that are the main responsible of the long-term radiotoxicity of nuclear waste. One of the key points that have been identified for the operation of an industrial-scale ADS is the need of continuously monitoring the reactivity of the subcritical system during operation. For this reason, since the 1990s a number of experiments have been conducted in zero-power subcritical assemblies (MUSE, RACE, KUCA, Yalina, GUINEVERE/FREYA) in order to experimentally validate these techniques. In this context, the present thesis is concerned with the validation of reactivity monitoring techniques at the Yalina-Booster subcritical assembly. This assembly belongs to the Joint Institute for Power and Nuclear Research (JIPNR-Sosny) of the National Academy of Sciences of Belarus. Experiments concerning reactivity monitoring have been performed in this facility under the EUROTRANS project of the 6th EU Framework Program in year 2008 under the direction of CIEMAT. Two types of experiments have been carried out: experiments with a pulsed neutron source (PNS) and experiments with a continuous source with short interruptions (beam trips). For the case of the first ones, PNS experiments, two fundamental techniques exist to measure the reactivity, known as the prompt-to-delayed neutron area-ratio technique (or Sjöstrand technique) and the prompt neutron decay constant technique. However, previous experiments have shown the need to apply correction techniques to take into account the spatial and energy effects present in a real system and thus obtain accurate values for the reactivity. In this thesis, these corrections have been investigated through simulations of the system with the Monte Carlo code MCNPX. This research has also served to propose a generalized version of these techniques where relationships between the reactivity of the system and the measured quantities are obtained through Monte Carlo simulations. The second type of experiments, with a continuous source with beam trips, is more likely to be employed in an industrial ADS. The generalized version of the techniques developed for the PNS experiments has also been applied to the result of these experiments. Furthermore, the work presented in this thesis is the first time, to my knowledge, that the reactivity of a subcritical system has been monitored during operation simultaneously with three different techniques: the current-to-flux, the source-jerk and the prompt neutron decay techniques. The cases analyzed include the fast variation of the system reactivity (insertion and extraction of a control rod) and the fast variation of the neutron source (long beam interruption and subsequent recovery).

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.

Feasibility assessment of the interactive use of a Monte Carlo algorithm in treatment planning for intraoperative electron radiation therapy

This work analysed the feasibility of using a fast, customized Monte Carlo (MC) method to perform accurate computation of dose distributions during pre- and intraplanning of intraoperative electron radiation therapy (IOERT) procedures. The MC method that was implemented, which has been integrated into a specific innovative simulation and planning tool, is able to simulate the fate of thousands of particles per second, and it was the aim of this work to determine the level of interactivity that could be achieved. The planning workflow enabled calibration of the imaging and treatment equipment, as well as manipulation of the surgical frame and insertion of the protection shields around the organs at risk and other beam modifiers. In this way, the multidisciplinary team involved in IOERT has all the tools necessary to perform complex MC dosage simulations adapted to their equipment in an efficient and transparent way. To assess the accuracy and reliability of this MC technique, dose distributions for a monoenergetic source were compared with those obtained using a general-purpose software package used widely in medical physics applications. Once accuracy of the underlying simulator was confirmed, a clinical accelerator was modelled and experimental measurements in water were conducted. A comparison was made with the output from the simulator to identify the conditions under which accurate dose estimations could be obtained in less than 3 min, which is the threshold imposed to allow for interactive use of the tool in treatment planning. Finally, a clinically relevant scenario, namely early-stage breast cancer treatment, was simulated with pre- and intraoperative volumes to verify that it was feasible to use the MC tool intraoperatively and to adjust dose delivery based on the simulation output, without compromising accuracy. The workflow provided a satisfactory model of the treatment head and the imaging system, enabling proper configuration of the treatment planning system and providing good accuracy in the dosage simulation.

Direct radiocarbon dates for Vindija G1 and Velika Pećina Late Pleistocene hominid remains

New accelerator mass spectrometry radiocarbon dates taken directly on human remains from the Late Pleistocene sites of Vindija and Velika Pećina in the Hrvatsko Zagorje of Croatia are presented. Hominid specimens from both sites have played critical roles in the development of current perspectives on modern human evolutionary emergence in Europe. Dates of ≈28 thousand years (ka) before the present (B.P.) and ≈29 ka B.P. for two specimens from Vindija G1 establish them as the most recent dated Neandertals in the Eurasian range of these archaic humans. The human frontal bone from Velika Pećina, generally considered one of the earliest representatives of modern humans in Europe, dated to ≈5 ka B.P., rendering it no longer pertinent to discussions of modern human origins. Apart from invalidating the only radiometrically based example of temporal overlap between late Neandertal and early modern human fossil remains from within any region of Europe, these dates raise the question of when early modern humans first dispersed into Europe and have implications for the nature and geographic patterning of biological and cultural interactions between these populations and the Neandertals.

Recombinant interleukin-1β dilates steelhead trout coronary microvessels : effect of temperature and role of the endothelium, nitric oxide and prostaglandins

© 2015. Published by The Company of Biologists Ltd. Acknowledgements We thank Wenjuan Xu and Xin Xu (Hein Lab) for their excellent instruction in microvessel techniques, Dr David Heeley (Biochemistry Department, MUN) for assistance with selecting an appropriate (non-vasoactive) protein stabilizer, Dr Zou (SFIRC, Aberdeen) for advice with regards to the use of rIL-1β and Gordon Nash (Gamperl Lab) for his assistance with the rIL-1β purification protocol. Funding This research was supported by a Natural Sciences and Engineering Research Council of Canada Discovery Grant [RGPIN249926] and Accelerator Supplement [RGPAS412325-2011] to A.K.G. a National Institutes of Health Grant [EY018420] to T.W.H., and a doctoral fellowship from Fundaçã o para a Ciência e a Tecnologia, Portugal [SFRH/BD/27497/2006] to I.A.S.F.C. Deposited in PMC for release after 12 months.