Optimization of the plasma start-up of the Seville spherical tokamak (SMART)

The SMall Aspect Ratio Tokamak (SMART) is currently being designed by the Plasma Science and Fusion Technology Group (PSFT) of the University of Seville. Therefore, several simulations and calculations regarding the di erent aspects of the reactor are needed. SMART is an experimental fusion reactor based on magnetic con nement, which has an aspect ratio (A) of >1.6, that classi es it as a spherical tokamak. Moreover, it has a central solenoid, which induces the toroidal electric eld to the plasma, two divertor coil sets and two poloidal eld coil sets. During this bachelor thesis simulations of the plasma startup in SMART has been carried out. A study optimizing the current of the solenoid and poloidal eld coils has been performed. Furthermore, an analysis of the suitable wall width for the outer, top and bottom walls studying the main parameters involved in the startup process such as the connection length (L) is conducted. Finally, the necessary electric eld to have a reliable breakdown is studied.El grupo "Plasma Science and Fusion Technology Group" (PSFT) de la Universidad de Sevilla está actualmente diseñando el reactor de fusión "SMall Aspect Ratio Tokamak" (SMART). Por lo tanto, se necesitan gran cantidad de simulaciones y cálculos sobre los diferentes aspectos del reactor. La relación de aspecto (A) en SMART es de >1,6, lo que lo clasifica como un tokamak esférico. Además, tiene un solenoide central, que induce el campo eléctrico toroidal al plasma, dos pares de bobinas divertoras y dos pares de anillos de campo poloidal. A lo largo de este trabajo de fin de grado se llevan a cabo distintas simulaciones del arranque del plasma en SMART. Se hace a su vez un estudio para optimizar la corriente de las bobinas del solenoide y del campo poloidal. Asimismo, se realiza un análisis de sobre cuál es el grosor adecuado para las paredes exteriores, superiores e inferiores, estudiando los principales parámetros que intervienen en el proceso de arranque, como la longitud de la conexión (L). Por último, se estudiará el campo eléctrico necesario para tener un "breakdown" que sea fiable.Universidad de Sevilla. Grado en Físic

Modelling for JET Vertical Stabilization System

Nuclear fusion is, in a sense, the opposite of nuclear fission. Fission, which is a mature technology, produces energy through the splitting of heavy atoms like uranium in controlled chain reactions. Unfortunately, the by-products of fission are highly radioactive and long lasting. On the other hand, fusion is the process by which the nuclei of two light atoms such as hydrogen are fused together to form a heavier (helium) nucleus, with energy produced as a by-product. Although controlled fusion is extremely technologically challenging, a fusion-power reactor would offer significant advantages over existing energy sources. This thesis is devoted to the control of tokamaks, magnetic confinement devices constructed in the shape of a torus (or doughnut). Tokamaks are the most promising of several proposed magnetic confinement devices. The need to improve the performance of modern tokamak operations has led to a further development of the plasma shape and position control systems. In particular, extremely elongated plasmas, with high vertical-instability growth rate, are envisaged to reach the required performance for ignition. This request for better performance from the experimentalists’ side has motivated the development of the new vertical-stabilization (VS) system at the JET tokamak, which has been proposed within the Plasma Control Upgrade project. This thesis presents the activity carried out to increase the capability of the VS system and to understand the operational limits in order to assess what can be done to improve the overall performance with the existing hardware and control system so as to minimize the impact on JET operation. The first objective of this work is the analysis of the new diagnostic system and the influence of the mechanical structure on the magnetic measurements used as diagnostics by the VS controller; the main focus is on the influence on the controller performance in the presence of large perturbations. The second objective is to design a new controlled variable to increase the performance of the VS system. The third objective is to provide an equivalent model of an ELM (Edge Localized Mode), in terms of internal plasma profile parameters via best fit of the vertical velocity estimation. The last objective is to obtain a reliable and accurate model of the overall system, based on the new platform MARTe, developed at JET and useful also for other devices

Dynamic analysis of the ITER Fast-Ion Loss Detector

La sociedad se enfrenta en nuestros días a la búsqueda de una fuente de energía limpia y fiable con la que superar la actual dependencia de los combustibles fósiles. En este sentido, la energía producida por el Sol ha despertado el interés de la comunidad científica. Desde mediados del siglo pasado, muchos han sido los avances en el campo de la fusión nuclear. No obstante, para demostrar su viabilidad técnica y económica es necesaria una apuesta sin precedentes. La Unión Europea, China, India, Japón, Corea del Sur, Rusia y los Estados Unidos cooperan en la construcción del reactor de fusión nuclear ITER, que va a suponer un auténtico hito en la historia de la humanidad. Desde un punto de vista ingenieril, una organización suficientemente robusta entre los distintos miembros es clave para garantizar un alto nivel de seguridad, siendo indispensable estructurar el trabajo y seguir un proceso de diseño estandarizado. Dentro de la estructura del proyecto se encuentra el sistema de diagnósticos. Estos diagnósticos juegan un papel esencial en la investigación de la fusión nuclear. En particular, uno de esos diagnósticos es el Detector de Pérdidas de Iones Rápidos (FILD), que proporciona información relevante acerca de la eficiencia de la reacción nuclear y del posible daño que se produce en las paredes del interior del reactor. A día de hoy, el diseño del FILD de ITER se encuentra aún en una fase conceptual, si bien encara la preparación para la revisión antes de pasar a la siguiente fase de diseño. Durante esa revisión, un equipo multidisciplinar de ITER evaluará el cumplimiento de los requerimientos del diseño conceptual propuesto. Para ello, es necesaria la entrega, en plazo y formas adecuados, de la documentación que lo justifica. El presente trabajo fin de grado ha tenido por objeto investigar y reunir todas las herramientas requeridas para realizar el análisis dinámico del FILD de ITER, cuyos resultados habrá que recoger debidamente en un informe entregable junto con el resto de la documentación exigida antes de la revisión del diseño. Puesto que el estado exigido para la revisión conceptual de ese informe es preliminar, no se ha buscado obtener unos resultados definitivos. Más bien, se ha pretendido entender a fondo las particularidades de este tipo de análisis, siguiendo las directrices propuestas por la Organización de ITER. En particular, el estudio dinámico del FILD de ITER se ha llevado a cabo mediante un análisis de espectro de respuesta, un método ampliamente utilizado en la industria nuclear por su sencillez y rapidez a la hora de obtener resultados conservadores. Aunque el análisis de espectro de respuesta se ha realizado con la ayuda del paquete ANSYS, ha sido crucial entender desde un punto de vista teórico el problema; en concreto, entender bien los sistemas de un grado de libertad, puesto que el estudio de los sitemas de N grados de libertad puede reducirse al estudio de N sistemas de un grado de libertad por medio de la técnica de superposición modal. Finalmente, se han obtenido unos resultados interesantes para un modelo de FILD ligeramente distinto, con un punto de apoyo que no estaría en la realidad, pero que ha sido necesario usarlo teniendo en cuenta la linealidad del análisis usado y el estado prematuro del diseño mecánico empleado.Nowadays, the society faces the search for a reliable environmentally-friendly energy source in order to overcome the current dependency on the fossil fuels. In this way, the energy released by the Sun has awakened interest in the scientific community. Since the middle of the 20th century, there have been a lot of breakthroughs in the field of nuclear fusion. However, an unprecedented investment is necessary to prove the technical and economic feasibility of fusion power. The EU, China, India, Japan, South Korea, the Russian Federation and the USA are collaborating in the construction of the ITER nuclear fusion reactor, which will mark a milestone in the history of humanity. From the point of view of engineering, a sufficiently robust organization among the different members is key to ensure a high safety level, being indispensable to structure the work and to follow a standardized design process. Inside the project structure, the diagnostics system is found. These diagnostics play an essential part in the nuclear fusion research. In particular, one of those diagnostics is the Fast-Ion Loss Detector (FILD), which provides important information about the nuclear reaction efficiency and the possible damage to the inner walls of the reactor. As of today, the ITER FILD design is still at a conceptual stage, but it faces the preparation for the previous review to go ahead until the next design phase. During this review, a multidisciplinary team from ITER will assess the compliance of the conceptual design proposed with the requirements. For that reason, it will be required to properly deliver the documentation that justifies it. The aim of the present final project has been to study and collect all the tools needed to perform a dynamic analysis of the ITER FILD, whose results will have to be appropiately included in a deliverable together with the rest of documentation demanded before the design review. Since the state that is required for the conceptual review is preliminar, the goal has not been to obtain definite results. On the contrary, the scope of this work has been limited by the well understanding of this type of analysis, following the guidelines proposed by ITER Organization. In particular, the dynamic study of ITER FILD has been carried out by means of a response spectrum analysis, a method widely used by the nuclear industry because of its simplicity and rapidity in obtaining conservative results. Even though the response spectrum analysis has been performed with the help of the ANSYS package, it has been crutial to understand the problem from a theoretical point of view; specifically, to well understand the single degree-of-freedom systems, since the study of the N degree-of-freedom systems can be reduced to the study of N single degree-of-freedom systems by using the modal superposition technique. Finally, some interesting results have been obtained for a slightly different FILD model, with a support point that would not exist in reality, but that has been necessary taking into consideration that the analysis used is linear by nature and because of the premature state of the mechanical design used.Universidad de Sevilla. Grado en Ingeniería Aeroespacia

Moving forward the thermal-hydraulic modeling of superconducting magnets for ITER: from the Insert Coils to the full magnets

L'abstract è presente nell'allegato / the abstract is in the attachmen

Development and Validation of a Computational Tool for Fusion Reactors\u27 System Analysis

On the roadmap to fusion energy the development and the operation of a demonstration power plant (DEMO) is the next step after ITER, a key facility currently devoted to the exploration of the physics aspects for self-sustained fusion plasmas with sizes and fusion power comparable to those attended in fusion power plants (FPP). Fusion systems codes are essential computational tools aimed to simulate the physics and the engineering features of a FPP. The main objective of a system code is to find one (or more) reactor configurations which simultaneously comply with physics operational limits, engineering constraints and net electric output requirements. As such simulation tools need to scope many design solutions over a large parameter phase space, they rely on rather basic physics and engineering models (mostly at zero or one-dimensional level) and on a relatively large number of input specifications. Within the conceptual design of a FPP, systems codes are interfaced to the detailed transport codes and engineering platforms, which operate in much larger time scales. To fill the gap between systems and the detailed transport and engineering codes the high-fidelity system/design tool MIRA (Modular Integrated Reactor Analysis) has been developed. MIRA relies on a modular structure and provides a refined FPP system analysis, with the primary goal of generating a more robust plant baseline. It incorporates into a unique computing environment a mathematical algorithm for the utmost tokamak fusion problems, including two-dimensional plasma magnetic equilibrium and core physics, transport of neutron and photon radiations emitted from the plasma and electromagnetic and engineering characterization of the toroidal field (TF) and poloidal field (PF) field coil systems. Most of the implemented modules rely on higher spatial resolution compared to presently available system codes, such as PROCESS. The multiphysics MIRA approach has been applied to the DEMO 2015 baseline, generated by means of the PROCESS system code. The analysis has been carried out by taking an identical set of input assumptions and requirements (e.g. same fusion power, major radius and aspect ratio) and observing the response on certain figures of merit. This verification study has featured the violation of some constraining conditions imposed on plasma safety factor, TF ripple and plasma burn time. The DEMO 2015 baseline has been found not in line with all the imposed requirements and constraints, hence necessitates a set of active measures on some of the input parameters. Such measures have been reported in form of parameter scans, where three variables have been identified, such as plasma internal inductance, blanket breeding zone inboard thickness and vacuum vessel/TF coil gap radial outboard width. The addressed sensitivity analyses have shown non-trivial inter-parametric dependencies, never explored in fusion system analyses. For instance, large influences of the plasma internal inductance on safety factor, plasma shape, density and temperature features, peak divertor flux and plasma burn time have been observed. Moreover, an optimal overall breeding blanket + TF coil inboard width has been observed with respect to the maximization of the plasma burn time, representing a meeting point between neutronic tritium breeding and technological limits in central solenoid and TF coils superconducting cables. These outcomes have inspired important changes in the way of designing a tokamak reactor like DEMO, where more extended analyses of the key physics and engineering aspects of the reactor can speed up and improve the design process of a FPP

Smart Energy, Plasma and Nuclear Systems

The extended papers in this Special Issue cover the topics of smart energy, nuclear systems, and micro energy grids. In “Electrical Loads and Power Systems for the DEMO Nuclear Fusion Project” and “Energy Analysis for the Connection of the Nuclear Reactor DEMO to the European Electrical Grid”, the authors introduce a European DEMO project. In “Comparison and Design of Resonant Network Considering the Characteristics of a Plasma Generator” the authors present a theoretical analysis and experimental study on the resonant network of the power conditioning system (PCS). In “Techno-Economic Evaluation of Interconnected Nuclear-Renewable Micro Hybrid Energy Systems with Combined Heat and Power”, the authors conducted a sensitivity analysis to identify the impact of the different variables on the investigated systems. In “Fault Current Tracing and Identification via Machine Learning Considering Distributed Energy Resources in Distribution Networks”, the authors propose a current tracing method to model the single distribution feeder as several independent parallel connected virtual lines, with the result of tracing the detailed contribution of different current sources to the power line current. From the five extended papers, we observe that the SEGE is actively engaged in smart grid and green energy techniques. We hope that the readers enjoy this Special Issue

DTT - Divertor Tokamak Test facility - Interim Design Report

The “Divertor Tokamak Test facility, DTT” is a milestone along the international program aimed at demonstrating – in the second half of this century – the feasibility of obtaining to commercial electricity from controlled thermonuclear fusion. DTT is a Tokamak conceived and designed in Italy with a broad international vision. The construction will be carried out in the ENEA Frascati site, mainly supported by national funds, complemented by EUROfusion and European incentive schemes for innovative investments. The project team includes more than 180 high-standard researchers from ENEA, CREATE, CNR, INFN, RFX and various universities. The volume, entitled DTT Interim Design Report (“Green Book” from the colour of the cover), briefly describes the status of the project, the planning of the design future activities and its organizational structure. The publication of the Green Book also provides an occasion for thorough discussions in the fusion community and a broad international collaboration on the DTT challenge