Application of gas-cooled Accelerator Driven System (ADS) transmutation devices to sustainable nuclear energy development

The conceptual design of a pebble bed gas-cooled transmutation device is shown with the aim to evaluate its potential for its deployment in the context of the sustainable nuclear energy development, which considers high temperature reactors for their operation in cogeneration mode, producing electricity, heat and Hydrogen. As differential characteristics our device operates in subcritical mode, driven by a neutron source activated by an accelerator that adds clear safety advantages and fuel flexibility opening the possibility to reduce the nuclear stockpile producing energy from actual LWR irradiated fuel with an efficiency of 45?46%, either in the form of Hydrogen, electricity, or both.

Introducing Nuclear Energy to High-School Students: The Spanish Young Generation in Nuclear (Jóvenes Nucleares) Lectures

One of the main goals of Spanish Young Generation (JJNN) is to spread knowledge about nuclear energy, not only pointing out its advantages and its role in our society, but also trying to correct some of the ideas that are due to the biased information and to the lack of knowledge. With this goal in mind, lectures were given in several high schools, aimed at students ranging from 14 to 18 years old. This paper explains the experience accumulated during those talks and the conclusions that can be drawn, so as to better focus the communication about nuclear energy, especially the one aimed at a young public. In order to evaluate the degree of knowledge and information on a specific topic of a given group of individuals, statistical methods must be used. At the beginning of each lecture (and sometimes at the end, in order to evaluate the impact of the talk) the students were submitted to a short survey conducted by Spanish Young Generation. It consisted in eight questions, dealing with the relation between the main environmental issues (global warming, acid rain, radioactive waste…) and nuclear energy. The answers can be surprising, especially for professionals of the nuclear field who, since they are so familiar with this topic, often forget that this is just the case of a minority of people. A better knowledge of the degree of information of a given group enables to focus and personalize the communication. Another communication tool is the direct contact with students: it starts with their questions, which can then lead to a small debate. If the surveys inform about the topics they are unaware of, the direct exchange with them enables to find the most effective way to provide them the information. Of course, it depends a lot on the public attending the talk (age, background…) and on the debate following the talk: a good communication, adapted to the public, is necessary. Therefore, the outcome of the performed exercise is that Spanish teenagers have still a lack of knowledge about nuclear energy. We can learn that items that are evident for nuclear young professionals are unknown for high school teenagers

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.

Aplicación de la metodología ISA al análisis de la secuencia de pérdida del sistema de agua de refrigeración de componentes en una central nuclear con reactor de agua a presión

La metodología Integrated Safety Analysis (ISA), desarrollada en el área de Modelación y Simulación (MOSI) del Consejo de Seguridad Nuclear (CSN), es un método de Análisis Integrado de Seguridad que está siendo evaluado y analizado mediante diversas aplicaciones impulsadas por el CSN; el análisis integrado de seguridad, combina las técnicas evolucionadas de los análisis de seguridad al uso: deterministas y probabilistas. Se considera adecuado para sustentar la Regulación Informada por el Riesgo (RIR), actual enfoque dado a la seguridad nuclear y que está siendo desarrollado y aplicado en todo el mundo. En este contexto se enmarcan, los proyectos Safety Margin Action Plan (SMAP) y Safety Margin Assessment Application (SM2A), impulsados por el Comité para la Seguridad de las Instalaciones Nucleares (CSNI) de la Agencia de la Energía Nuclear (NEA) de la Organización para la Cooperación y el Desarrollo Económicos (OCDE) en el desarrollo del enfoque adecuado para el uso de las metodologías integradas en la evaluación del cambio en los márgenes de seguridad debidos a cambios en las condiciones de las centrales nucleares. El comité constituye un foro para el intercambio de información técnica y de colaboración entre las organizaciones miembro, que aportan sus propias ideas en investigación, desarrollo e ingeniería. La propuesta del CSN es la aplicación de la metodología ISA, especialmente adecuada para el análisis según el enfoque desarrollado en el proyecto SMAP que pretende obtener los valores best-estimate con incertidumbre de las variables de seguridad que son comparadas con los límites de seguridad, para obtener la frecuencia con la que éstos límites son superados. La ventaja que ofrece la ISA es que permite el análisis selectivo y discreto de los rangos de los parámetros inciertos que tienen mayor influencia en la superación de los límites de seguridad, o frecuencia de excedencia del límite, permitiendo así evaluar los cambios producidos por variaciones en el diseño u operación de la central que serían imperceptibles o complicados de cuantificar con otro tipo de metodologías. La ISA se engloba dentro de las metodologías de APS dinámico discreto que utilizan la generación de árboles de sucesos dinámicos (DET) y se basa en la Theory of Stimulated Dynamics (TSD), teoría de fiabilidad dinámica simplificada que permite la cuantificación del riesgo de cada una de las secuencias. Con la ISA se modelan y simulan todas las interacciones relevantes en una central: diseño, condiciones de operación, mantenimiento, actuaciones de los operadores, eventos estocásticos, etc. Por ello requiere la integración de códigos de: simulación termohidráulica y procedimientos de operación; delineación de árboles de sucesos; cuantificación de árboles de fallos y sucesos; tratamiento de incertidumbres e integración del riesgo. La tesis contiene la aplicación de la metodología ISA al análisis integrado del suceso iniciador de la pérdida del sistema de refrigeración de componentes (CCWS) que genera secuencias de pérdida de refrigerante del reactor a través de los sellos de las bombas principales del circuito de refrigerante del reactor (SLOCA). Se utiliza para probar el cambio en los márgenes, con respecto al límite de la máxima temperatura de pico de vaina (1477 K), que sería posible en virtud de un potencial aumento de potencia del 10 % en el reactor de agua a presión de la C.N. Zion. El trabajo realizado para la consecución de la tesis, fruto de la colaboración de la Escuela Técnica Superior de Ingenieros de Minas y Energía y la empresa de soluciones tecnológicas Ekergy Software S.L. (NFQ Solutions) con el área MOSI del CSN, ha sido la base para la contribución del CSN en el ejercicio SM2A. Este ejercicio ha sido utilizado como evaluación del desarrollo de algunas de las ideas, sugerencias, y los algoritmos detrás de la metodología ISA. Como resultado se ha obtenido un ligero aumento de la frecuencia de excedencia del daño (DEF) provocado por el aumento de potencia. Este resultado demuestra la viabilidad de la metodología ISA para obtener medidas de las variaciones en los márgenes de seguridad que han sido provocadas por modificaciones en la planta. También se ha mostrado que es especialmente adecuada en escenarios donde los eventos estocásticos o las actuaciones de recuperación o mitigación de los operadores pueden tener un papel relevante en el riesgo. Los resultados obtenidos no tienen validez más allá de la de mostrar la viabilidad de la metodología ISA. La central nuclear en la que se aplica el estudio está clausurada y la información relativa a sus análisis de seguridad es deficiente, por lo que han sido necesarias asunciones sin comprobación o aproximaciones basadas en estudios genéricos o de otras plantas. Se han establecido tres fases en el proceso de análisis: primero, obtención del árbol de sucesos dinámico de referencia; segundo, análisis de incertidumbres y obtención de los dominios de daño; y tercero, cuantificación del riesgo. Se han mostrado diversas aplicaciones de la metodología y ventajas que presenta frente al APS clásico. También se ha contribuido al desarrollo del prototipo de herramienta para la aplicación de la metodología ISA (SCAIS). ABSTRACT The Integrated Safety Analysis methodology (ISA), developed by the Consejo de Seguridad Nuclear (CSN), is being assessed in various applications encouraged by CSN. An Integrated Safety Analysis merges the evolved techniques of the usually applied safety analysis methodologies; deterministic and probabilistic. It is considered as a suitable tool for assessing risk in a Risk Informed Regulation framework, the approach under development that is being adopted on Nuclear Safety around the world. In this policy framework, the projects Safety Margin Action Plan (SMAP) and Safety Margin Assessment Application (SM2A), set up by the Committee on the Safety of Nuclear Installations (CSNI) of the Nuclear Energy Agency within the Organization for Economic Co-operation and Development (OECD), were aimed to obtain a methodology and its application for the integration of risk and safety margins in the assessment of the changes to the overall safety as a result of changes in the nuclear plant condition. The committee provides a forum for the exchange of technical information and cooperation among member organizations which contribute their respective approaches in research, development and engineering. The ISA methodology, proposed by CSN, specially fits with the SMAP approach that aims at obtaining Best Estimate Plus Uncertainty values of the safety variables to be compared with the safety limits. This makes it possible to obtain the exceedance frequencies of the safety limit. The ISA has the advantage over other methods of allowing the specific and discrete evaluation of the most influential uncertain parameters in the limit exceedance frequency. In this way the changes due to design or operation variation, imperceptibles or complicated to by quantified by other methods, are correctly evaluated. The ISA methodology is one of the discrete methodologies of the Dynamic PSA framework that uses the generation of dynamic event trees (DET). It is based on the Theory of Stimulated Dynamics (TSD), a simplified version of the theory of Probabilistic Dynamics that allows the risk quantification. The ISA models and simulates all the important interactions in a Nuclear Power Plant; design, operating conditions, maintenance, human actuations, stochastic events, etc. In order to that, it requires the integration of codes to obtain: Thermohydraulic and human actuations; Even trees delineation; Fault Trees and Event Trees quantification; Uncertainty analysis and risk assessment. This written dissertation narrates the application of the ISA methodology to the initiating event of the Loss of the Component Cooling System (CCWS) generating sequences of loss of reactor coolant through the seals of the reactor coolant pump (SLOCA). It is used to test the change in margins with respect to the maximum clad temperature limit (1477 K) that would be possible under a potential 10 % power up-rate effected in the pressurized water reactor of Zion NPP. The work done to achieve the thesis, fruit of the collaborative agreement of the School of Mining and Energy Engineering and the company of technological solutions Ekergy Software S.L. (NFQ Solutions) with de specialized modeling and simulation branch of the CSN, has been the basis for the contribution of the CSN in the exercise SM2A. This exercise has been used as an assessment of the development of some of the ideas, suggestions, and algorithms behind the ISA methodology. It has been obtained a slight increase in the Damage Exceedance Frequency (DEF) caused by the power up-rate. This result shows that ISA methodology allows quantifying the safety margin change when design modifications are performed in a NPP and is specially suitable for scenarios where stochastic events or human responses have an important role to prevent or mitigate the accidental consequences and the total risk. The results do not have any validity out of showing the viability of the methodology ISA. Zion NPP was retired and information of its safety analysis is scarce, so assumptions without verification or approximations based on generic studies have been required. Three phases are established in the analysis process: first, obtaining the reference dynamic event tree; second, uncertainty analysis and obtaining the damage domains; third, risk quantification. There have been shown various applications of the methodology and advantages over the classical PSA. It has also contributed to the development of the prototype tool for the implementation of the ISA methodology (SCAIS).

Potential common radiation problems for components and diagnostics in future magnetic and inertial confinement fusion devices

This work aims at identifying commonpotentialproblems that futurefusiondevices will encounter for both magnetic and inertialconfinement approaches in order to promote joint efforts and to avoid duplication of research. Firstly, a comparison of radiation environments found in both fusion reaction chambers will be presented. Then, wall materials, optical components, cables and electronics will be discussed, pointing to possible future areas of common research. Finally, a brief discussion of experimental techniques available to simulate the radiation effect on materials is included

Temperature Evolution and Light Species Diffusion in Armor and Structural Material for Inertial Fusion Reactor Chambers: a Case for HiPER 4a

One of the most advance designs for HiPER fusion reactor is a spherical chamber 10 m in diameter based on dry wall concept. In this system, the first wall will have to withstand short energy pulses of 5 to 20 MJ at a repetition rate of 0.5-10 Hz mostly in form of X-rays and charged particles. To avoid melting of the inner surface, the first wall consists on a thin armor attached to the structural material. Thickness (th) and material of each layer have to be chosen to assure the proper functioning of the facility during its planned lifetime.

Molecular dynamics simulations of lead and lithium in liquid phase

Pb17Li is today a reference breeder material in diverse fusion R&D programs worldwide. Extracting dynamic and structural properties of liquid LiPb mixtures via molecular dynamics simulations, represent a crucial step for multiscale modeling efforts in order to understand the suitability of this compound for future Nuclear Fusion technologies. At present a Li-Pb cross potential is not available in the literature. Here we present our first results on the validation of two semi-empirical potentials for Li and Pb in liquid phase. Our results represent the establishment of a solid base as a previous but crucial step to implement a LiPb cross potential. Structural and thermodynamical analyses confirm that the implemented potentials for Li and Pb are realistic to simulate both elements in the liquid phase.

Influence of termal fluctuations in radiation damage cascade production and defect dynamics in tungsten

The detailed study of the deterioration suffered by the materials of the components of a nuclear facility, in particular those forming part of the reactor core, is a topic of great interest which importance derives in large technological and economic implications. Since changes in the atomic-structural properties of relevant components pose a risk to the smooth operation with clear consequences for security and life of the plant, controlling these factors is essential in any development of engineering design and implementation. In recent times, tungsten has been proposed as a structural material based on its good resistance to radiation, but still needs to be done an extensive study on the influence of temperature on the behavior of this material under radiation damage. This work aims to contribute in this regard. Molecular Dynamics (MD) simulations were carried out to determine the influence of temperature fluctuations on radiation damage production and evolution in Tungsten. We have particularly focused our study in the dynamics of defect creation, recombination, and diffusion properties. PKA energies were sampled in a range from 5 to 50 KeV. Three different temperature scenarios were analyzed, from very low temperatures (0-200K), up to high temperature conditions (300-500 K). We studied the creation of defects, vacancies and interstitials, recombination rates, diffusion properties, cluster formation, their size and evolution. Simulations were performed using Lammps and the Zhou EAM potential for W

Assessment of CTF boiling transition and critical heat flux modeling capabilities using the OECD/NRC BFBT and PSBT benchmark databases

The need to refine models for best-estimate calculations, based on good-quality experimental data, has been expressed in many recent meetings in the field of nuclear applications. The modeling needs arising in this respect should not be limited to the currently available macroscopic methods but should be extended to next-generation analysis techniques that focus on more microscopic processes. One of the most valuable databases identified for the thermalhydraulics modeling was developed by the Nuclear Power Engineering Corporation (NUPEC), Japan. From 1987 to 1995, NUPEC performed steady-state and transient critical power and departure from nucleate boiling (DNB) test series based on the equivalent full-size mock-ups. Considering the reliability not only of the measured data, but also other relevant parameters such as the system pressure, inlet sub-cooling and rod surface temperature, these test series supplied the first substantial database for the development of truly mechanistic and consistent models for boiling transition and critical heat flux. Over the last few years the Pennsylvania State University (PSU) under the sponsorship of the U.S. Nuclear Regulatory Commission (NRC) has prepared, organized, conducted and summarized the OECD/NRC Full-size Fine-mesh Bundle Tests (BFBT) Benchmark. The international benchmark activities have been conducted in cooperation with the Nuclear Energy Agency/Organization for Economic Co-operation and Development (NEA/OECD) and Japan Nuclear Energy Safety (JNES) organization, Japan. Consequently, the JNES has made available the Boiling Water Reactor (BWR) NUPEC database for the purposes of the benchmark. Based on the success of the OECD/NRC BFBT benchmark the JNES has decided to release also the data based on the NUPEC Pressurized Water Reactor (PWR) subchannel and bundle tests for another follow-up international benchmark entitled OECD/NRC PWR Subchannel and Bundle Tests (PSBT) benchmark. This paper presents an application of the joint Penn State University/Technical University of Madrid (UPM) version of the well-known subchannel code COBRA-TF, namely CTF, to the critical power and departure from nucleate boiling (DNB) exercises of the OECD/NRC BFBT and PSBT benchmarks

From Secondary School To University: Attracting Young Students Towards A Career In Nuclear

From its creation, Spanish Young Generation in Nuclear (Jóvenes Nucleares, JJNN), a non-profit organization that depends on the Spanish Nuclear Society (SNE), has as an important scope to help spread knowledge about nuclear energy, not only pointing out its advantages and its role in our society, but also trying to correct some of the ideas that are due to the biased information and to the lack of knowledge. To try to have success in that goal, some high school lectures were taught and it has been organized regularly a Basic Course on Nuclear Science and Technology

Motivating Young Students to Start a Career in Nuclear: The Basic Course of Science and Nuclear Technology, An Educational Activity of Spanish Young Generation in Nuclear (JÓVENES NUCLEARES)

The main objective of this course, conducted by Jóvenes Nucleares (Spanish Young Generation in Nuclear, JJNN), a non-profit organization that depends on the Spanish Nuclear Society (SNE) is to pass on basic knowledge about Science and Nuclear Technology to the general public, mostly students and introduce them to its most relevant points. The purposes of this course are to provide general information, to answer the most common questions about Nuclear Energy and to motivate the young students to start a career in nuclear. Therefore, it is directed mainly to high school and university students, but also to general people that wants to learn about the key issues of such an important matter in our society. Anybody could attend the course, as no specific scientific education is required. The course is done at least once a year, during the Annual Meeting of the Spanish Nuclear Society, which takes place in a different Spanish city each time. The course is done also to whichever university or institution that asks for it to JJNN, with the only limit of the presenter´s availability. The course is divided into the following chapters: Physical nuclear and radiation principles, Nuclear power plants, Nuclear safety, Nuclear fuel, Radioactive waste, Decommission of nuclear facilities, Future nuclear power plants, Other uses of nuclear technology, Nuclear energy, climate change and sustainable development. The course is divided into 15 minutes lessons on the above topics, imparted by young professionals, experts in the field that belongs either to the Spanish Young Generation in Nuclear, either to companies and institutions related with nuclear energy. At the end of the course, a 200 pages book with the contents of the course is handed to every member of the audience. This book is also distributed in other course editions at high schools and universities in order to promote the scientific dissemination of the Nuclear Technology. As an extra motivation, JJNN delivers a course certificate to the assistants. At the end of the last edition course, in Santiago de Compostela, the assistants were asked to provide a feedback about it. Some really interesting lessons were learned, that will be very useful to improve next editions of the course. As a general conclusion of the courses it can be said that many of the students that have assisted to the course have increased their motivation in the nuclear field, and hopefully it will help the young talents to choose the nuclear field to develop their careers

The Key Role Transferring Knowledge of Nuclear España, The Spanish Nuclear Society Magazine

The magazine of the Spanish Nuclear Society (SNE), “Nuclear España” is a scientific-technical publication with almost thirty years of uninterrupted edition and more than 300 numbers published. Their pages approach technical subjects related to the nuclear energy, as well as the activities developed by the SNE, especially in national and international meetings. The main part of the magazine is composed by articles written by known specialist of the energy industry. One of the top goals of the magazine is to help on transferring the knowledge from the generation that built the nuclear power plants in Spain and the new generation of professionals that have started its nuclear career in the last years. Each number is monographic, trying to cover as many aspects on an issue as it is possible, with collaborations from the companies, the research centers and universities that helps to have complementary points of view. On the other hand the articles help to deep in the issue´s topic, broadening the view of the readers about the nuclear field and helping to share knowledge across the industry. The news section of the Magazine picks up the actuality of the sector as a whole. The editorial section reflects the opinion of the SNE Governing Board and the Magazine Committee on the subjects of interest in this field. On the other hand, the monthly interview sets out the professional outstanding opinions. With a total of eleven numbers per year, three of them have a noticeable international character: the one dedicated to the operative experiences on the Spanish and European nuclear power plants, the monographic issue devoted tothe Annual Meeting of the SNE and the international issue, which covers the last activities of the Spanish industry in international projects. Both first are bilingual issues (Spanish-English), whereas the international edition is published completely in English. Besides its diffusion through all the members of the SNE, the Magazine is distributed, in the national scope, to companies and organisms related to the nuclear power, universities, research centers, representatives of the Central, Autonomic and Local Administrations, mass media and communication professionals. It is also sent to the utilities and research centers in Europe, United States, South America and Asia.

Performance of a transmutation advanced device for sustainable energy application

Preliminary studies have been performed to design a device for nuclear waste transmutation and hydrogen generation based on a gas-cooled pebble bed accelerator driven system, TADSEA (Transmutation Advanced Device for Sustainable Energy Application). In previous studies we have addressed the viability of an ADS Transmutation device that uses as fuel wastes from the existing LWR power plants, encapsulated in graphite in the form of pebble beds, cooled by helium which enables high temperatures (in the order of 1200 K), to generate hydrogen from water either by high temperature electrolysis or by thermochemical cycles. For designing this device several configurations were studied, including several reflectors thickness, to achieve the desired parameters, the transmutation of nuclear waste and the production of 100 MW of thermal power. In this paper new studies performed on deep burn in-core fuel management strategy for LWR waste are presented. The fuel cycle on TADSEA device has been analyzed based on both: driven and transmutation fuel that had been proposed by the General Atomic design of a gas turbine-modular helium reactor. The transmutation results of the three fuel management strategies, using driven, transmutation and standard LWR spent fuel were compared, and several parameters describing the neutron performance of TADSEA nuclear core as the fuel and moderator temperature reactivity coefficients and transmutation chain, are also presented

Development of an adequate model for verification of design safety-margins of the HTTR nuclear test facility

This work is based on the prototype High Engineering Test Reactor (HTTR) of the Japan Agency of Energy Atomic (JAEA). Its objective is to describe an adequate deterministic model to be used in the assessment of its design safety margins via damage domains. The concept of damage domain is defined and it is shown its relevance in the ongoing effort to apply dynamic risk assessment methods and tools based on the Theory of Stimulated Dynamics (TSD). To illustrate, we present results of an abnormal control rod (CR) withdrawal during subcritical condition and its comparison with results obtained by JAEA. No attempt is made yet to actually assess the detailed scenarios, rather to show how the approach may handle events of its kind

The seminar of nuclear safety in advanced reactors and the seminar of nuclear fusion: two formative initiatives from Universidad Politécnica de Madrid and spanish young Generation in Nuclear (Jóvenes Nucleares)

panish Young Generation in Nuclear (Jóvenes Nucleares) is a commission of the Spanish Nuclear Society (SNE), whose main goals are to spread knowledge about nuclear energy among the society. Following this motivation, two Seminars have been carried out with the collaboration of the Technical University of Madrid: The Seminar of Nuclear Safety in Advanced Reactors (SRA) and the Seminar of Nuclear Fusion (SFN). The first one, which has been celebrated every year since 2010, aims to show clearly the advances that have been obtained in the section of safety with the new reactors, from a technical but simple point of view and without needing great previous nuclear engineering knowledge. The second one, which first edition was held in 2011, aims to give a general overview of the past, present and future situation of nuclear fusion technology, and was born as a result of the increasing interest of our Spanish Young Generation members in this technology.