Garching bei München, Germany
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Lampasi A.,ENEA | Zito P.,ENEA | Starace F.,ENEA | Costa P.,ENEA | And 6 more authors.
Fusion Engineering and Design | Year: 2017

This paper presents the design criteria and the preliminary characteristics of the power supply and electrical systems of the Divertor Tokamak Test (DTT) facility. The power supply system has to feed: 6 superconducting modules of the central solenoid, 6 poloidal field superconducting coils, 18 toroidal field superconducting coils designed for a current up to 50. kA, some coils for plasma fast control and vertical stabilization, the electron (ECRH) and ion (ICRH) cyclotron additional heating systems designed to deliver about 25. MW to the plasma, further 20. MW to the plasma generated by a neutral beam injector (NBI) and all the auxiliary systems and services.The analysis was carried out on a reference scenario with a plasma current of 6 MA, mainly to estimate the electrical power needed to operate the facility, but also to identify some design choices and component ratings. © 2017 The Authors.


Bachmann C.,EUROfusion Consortium | Biel W.,Jülich Research Center | Ciattaglia S.,EUROfusion Consortium | Federici G.,EUROfusion Consortium | And 5 more authors.
Fusion Engineering and Design | Year: 2017

An essential goal of the EU fusion roadmap is the development of design and technology of a Demonstration Fusion Power Reactor (DEMO) to follow ITER. A pragmatic approach is advocated considering a pulsed tokamak based on mature technologies and reliable regimes of operation, extrapolated as far as possible from the ITER experience. The EUROfusion Power Plant Physics and Technology Department (PPPT) started the conceptual design of DEMO in 2014, see Federici et al. (2014) .This article defines, based on ASME III, the categories of loads to be considered in the design of the DEMO components, defines the categorization of load conditions based on their expected occurrence and provides the correlation of acceptable component damage levels. It furthermore defines the load combinations to be considered in the conceptual design phase of DEMO. Furthermore, with exception of heat loads from plasma particles and radiation to the plasma facing components, the most important load cases are described and quantified. These include (i) electromagnetic (EM) loads due to toroidal field coil fast discharge, (ii) EM loads in fast and slow plasma disruptions due to eddy and halo currents, (iii) seismic loads, and (vi) pressure loads in the dominant incident/accident events. © 2017 The Author(s).


Bruzzone P.,Ecole Polytechnique Federale de Lausanne | Sedlak K.,Ecole Polytechnique Federale de Lausanne | Stepanov B.,Ecole Polytechnique Federale de Lausanne | Wesche R.,Ecole Polytechnique Federale de Lausanne | And 4 more authors.
IEEE Transactions on Applied Superconductivity | Year: 2016

The Toroidal field coils (TFC) of the EUROFusion DEMO reactor call for Nb3Sn conductor with high field and high current. Another major requirement is cost-effectiveness, to keep the ratio of investment to electric power in the same range of the competing energy sources (fission, hydro, coal, gas, etc.). The TFC proposed by the Swiss Plasma Center (SPC) is based on a double-layer Nb3Sn/NbTi winding. A react-and-wind flat cable is the core of the Nb3Sn conductor, with six grades to minimize the cost and maintain a roughly constant temperature margin of 1.5 K over the winding cross section. A short length section of the high grade Nb3Sn conductor has been manufactured using relevant industrial cabling equipment. One hundred kilograms of 1.5-mm Nb3Sn strand has been procured at WST with average Jc up to 15% higher than specified (Jc≥ 1000A/mm2 at 12 T/4.2 K). A dedicated cabling line has been set up at TRATOS cavi (Italy), producing over 350 m of dummy cable and about 13 m of superconducting cable. The assembly of the cable into a conduit by longitudinal laser welding of two steel profiles is demonstrated, including the QA procedures. A test sample has been prepared at SPC by heat treating straight sections of the cable and encasing it into a steel jacket after the heat treatment to minimize the thermal strain. The test was carried out in three test campaigns at the EDIPO facility at SPC. The test program includes Dc performance at the relevant operating conditions. An assessment of the conductor test results in terms of strand performance suggests that the applicable thermal strain is less than -0.33%. The performance is stable upon load cycles. © 2002-2011 IEEE.


Wesche R.,Ecole Polytechnique Federale de Lausanne | Sedlak K.,Ecole Polytechnique Federale de Lausanne | Bykovsky N.,Ecole Polytechnique Federale de Lausanne | Bruzzone P.,Ecole Polytechnique Federale de Lausanne | And 2 more authors.
IEEE Transactions on Applied Superconductivity | Year: 2016

The design of the European DEMO, i.e., the future fusion tokamak planned after ITER, is being developed under the coordination of the EUROfusion Consortium. This paper reports the design optimization of the toroidal field (TF) winding pack and its corresponding react-and-wind conductor, and a new design study of the central solenoid (CS). The optimization of the TF coil is driven by the results of the mechanical analysis that revealed an unacceptable stress accumulation in some locations of the previous proposal of the TF winding pack. The design study of the CS coil is done with the aim of minimizing the outer radius, while maintaining the magnetic flux defined in the PROCESS system code. The results of the design study, namely the optimized CS coil radius, current density, hoop stress, and the field map, define the initial information that will be needed in future for designing the DEMO CS winding pack and conductor. Contrary to the former similar studies, no upper limit is set for the peak field of the CS, implicitly allowing the use of high-temperature superconductors wherever the current density of Nb3Sn at the operating field is too low. © 2002-2011 IEEE.


Fischer U.,Karlsruhe Institute of Technology | Bachmann C.,EUROfusion Consortium | Palermo I.,CIEMAT | Pereslavtsev P.,Karlsruhe Institute of Technology | Villari R.,ENEA
Fusion Engineering and Design | Year: 2015

This paper addresses the neutronic requirements a DEMO fusion power plant needs to fulfil for a reliable and safe operation. The major requirement is to ensure Tritium self-sufficiency taking into account the various uncertainties and plant-internal losses that occur during DEMO operation. A further major requirement is to ensure sufficient protection of the superconducting magnets against the radiation penetrating in-vessel components and vessel. Reliable criteria for the radiation loads need to be defined and verified to ensure the reliable operation of the magnets over the lifetime of DEMO. Other issues include radiation induced effects on structural materials such as the accumulated displacement damage, the generation of gases such as helium which may deteriorate the material performance. The paper discusses these issues and their impact on design options for DEMO taking into account results obtained in the frame of European Power Plant Physics and Technology (PPPT) 2013 programme activities with DEMO models employing the helium cooled pebble bed (HCPB), the helium cooled lithium lead (HCLL), and the water-cooled (WCLL) blanket concepts. © 2015.


Li M.,Max Planck Institute for Plasma Physics (Garching) | Maviglia F.,EUROfusion Consortium | Federici G.,EUROfusion Consortium | You J.-H.,Max Planck Institute for Plasma Physics (Garching)
Fusion Engineering and Design | Year: 2016

One possibility to mitigate the maximum high heat flux (HHF) load on the target is to sweep the position of the strike-point back and forth periodically in order to spread the peak thermal load over a wider width. The aim of this work is to investigate the thermal and structure-mechanical responses of a water-cooled tungsten mono-block target under cyclic HHF loads which are applied in sweeping modes. The study was performed by means of finite element analysis using an ITER-like target geometry. Extensive parametric simulations were carried out for a wide range of HHF loads and for selected sweeping amplitudes and frequencies, respectively. The simulation shows that the maximum temperature and the maximum heat flux to the coolant can be significantly reduced by sweeping. For the parameters studied in this work (0.5-4 Hz, 5-20 cm), higher sweeping frequency or smaller sweeping amplitude offers advantages in terms of fatigue lifetime of interlayer. Sweeping is suitable for the stationary loading if the sweeping frequency is high enough (e.g. 4 Hz) based on the fatigue lifetime prediction of interlayer. © 2015 EURATOM. Published by Elsevier B.V. All rights reserved.


Meszaros B.,EUROfusion Consortium | Shannon M.,EUROfusion Consortium | Marzullo D.,University of Naples Federico II | Woodley C.,Culham Center for Fusion Energy | And 2 more authors.
Fusion Engineering and Design | Year: 2015

The EUROfusion Consortium is setting up - as part of the EU Fusion Roadmap - the framework for the implementation of the (pre)conceptual design phase of the DEMO reactor. Configuration management needs have been identified as one of the key elements of this framework and is the topic of this paper, in particular the configuration of the CAD design data. The desire is to keep the definition and layout of the corresponding systems "light weight" and relatively easy to manage, whilst simultaneously providing a level of detail in the definition of the design configuration that is fit for the purpose of a conceptual design. This paper aims to describe the steps followed during the definition of the configuration management system of the DEMO design data in terms of (i) the identification of the appropriate product data management system, (ii) the description of the philosophy of the configuration management of the design data, and (iii) the introduction of the most important enabling processes. © 2015 Elsevier B.V.


Coleman M.,EUROfusion Consortium | Coleman M.,Culham Center for Fusion Energy | Maviglia F.,EUROfusion Consortium | Bachmann C.,EUROfusion Consortium | And 6 more authors.
Fusion Engineering and Design | Year: 2016

One of the difficulties inherent in designing a future fusion reactor is dealing with uncertainty. As the major step between ITER and the commercial exploitation of nuclear fusion energy, DEMO will have to address many challenges - the natures of which are still not fully known. Unlike fission reactors, fusion reactors suffer from the intrinsic complexity of the tokamak (numerous interdependent system parameters) and from the dependence of plasma physics on scale - prohibiting design exploration founded on incremental progression and small-scale experimentation. For DEMO, this means that significant technical uncertainties will exist for some time to come, and a systems engineering design exploration approach must be developed to explore the reactor architecture when faced with these uncertainties. Important uncertainties in the context of fusion reactor design are discussed and a strategy for dealing with these is presented, treating the uncertainty in the first wall loads as an example. © 2016.


Froio A.,Polytechnic University of Turin | Bachmann C.,EUROfusion Consortium | Cismondi F.,EUROfusion Consortium | Savoldi L.,Polytechnic University of Turin | Zanino R.,Polytechnic University of Turin
Progress in Nuclear Energy | Year: 2016

A global, system-level thermal-hydraulic model of the EU DEMO tokamak fusion reactor is currently under development and implementation in a suitable software at Politecnico di Torino, including the relevant heat transfer and fluid dynamics phenomena, which affect the performance of the different cooling circuits and components and their integration in a consistent design. The model is based on an object-oriented approach using the Modelica language, which easily allows to preserve the high modularity required at this stage of the design. The first module of the global model will simulate the blanket cooling system and will be able to investigate different coolant options and different cooling schemes, to be adapted to the different blanket systems currently under development in the Breeding Blanket (BB) project. The paper presents the Helium-Cooled Pebble Bed (HCPB) module of the EU DEMO blanket cooling loops system model. The model is used to compare different schemes for the cooling of the different components of the HCPB BB, and to suggest improvements aimed at optimizing the pumping power required by the cooling system. The model is then used to analyse a pulsed scenario, characteristic of the EU DEMO operation. © 2016 EURATOM


Flammini D.,ENEA | Villari R.,ENEA | Moro F.,ENEA | Pizzuto A.,ENEA | Bachmann C.,EUROfusion Consortium
Fusion Engineering and Design | Year: 2016

The DEMO vacuum vessel, a massive water cooled double-walled steel vessel, is located behind breeding blankets and manifolds and it will be subjected to an intense neutron and photon irradiation. Therefore, a proper evaluation of the vessel nuclear heat loads is required to assure adequate cooling and, given the significant lifetime neutron fluence of DEMO, the radiation damage limit of the vessel needs to be carefully controlled. In the present work nuclear heating, radiation damage (DPA), helium production, neutron and photon fluxes have been calculated on the vacuum vessel at the inboard by means of MCNP5 using a 3D Helium Cooled Lithium Lead (HCLL) DEMO model with 1572MW of fusion power. In particular, the effect of the poloidal gap between the breeding-blanket segments on vacuum vessel nuclear loads has been estimated varying the gap width from 0 to 5cm. High values of the nuclear heating (≈1W/cm3), which might cause intense thermal stresses, were obtained in inboard equatorial zone. The effect of the poloidal gap on the nuclear heating resulted to be moderate (within 30%). The radiation damage limit of 2.75DPA on the vessel is almost met with 1cm of poloidal gap over DEMO lifetime. A comparison with Helium Cooled Pebble Bed blanket is also provided. © 2016 Davide Flammini.

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