Fusion for Energy F4E
Fusion for Energy F4E
Ezeberry J.,Nuclear Services |
Combescure D.,Fusion for Energy F4E
Nuclear Engineering and Design | Year: 2017
A suitable methodology to find the seismic floor response spectra (FRS) at a particular location of a structural system by a direct spectra-to-spectra method is presented in this work. This methodology was developed to obtain FRS inside the ITER Tokamak Complex building and the mechanical devices located in the building, as the Fusion Reactor Machine. Based on the well-established tools developed over the last decades in random vibration theory, the methodology developed totally in the frequency domain, has three main steps: First, the generation of a free-field acceleration PSD functions compatible with the Design Response Spectra; Second, the generation of the absolute acceleration PSD function at the locations of interest within the structure; Third, the determination of the FRS. To demonstrate the feasibility of the method, the FRS obtained at a different points of the Tokamak Complex are compared with the FRS obtained by means of the classical methods, based on time-histories of absolute acceleration obtained in time-domain. The comparison reveals a great accuracy of the proposed method. © 2017.
Farina D.,ENEA |
Henderson M.,ITER Organization |
Figini L.,ENEA |
Saibene G.,Fusion for Energy F4E
Physics of Plasmas | Year: 2014
The design of the ITER Electron Cyclotron Heating and Current Drive (EC H&CD) system has evolved in the last years both in goals and functionalities by considering an expanded range of applications. A large effort has been devoted to a better integration of the equatorial and the upper launchers, both from the point of view of the performance and of the design impact on the engineering constraints. However, from the analysis of the ECCD performance in two references H-mode scenarios at burn (the inductive H-mode and the advanced non-inductive scenario), it was clear that the EC power deposition was not optimal for steady-state applications in the plasma region around mid radius. An optimization study of the equatorial launcher is presented here aiming at removing this limitation of the EC system capabilities. Changing the steering of the equatorial launcher from toroidal to poloidal ensures EC power deposition out to the normalized toroidal radius ρ 0.6, and nearly doubles the EC driven current around mid radius, without significant performance degradation in the core plasma region. In addition to the improved performance, the proposed design change is able to relax some engineering design constraints on both launchers. © 2014 EURATOM.
Evans D.,Advanced Cryogenic Materials Ltd |
Knaster J.,ITER Organization |
Rajainmaki H.,Fusion for Energy F4E
IEEE Transactions on Applied Superconductivity | Year: 2012
The technology associated with the design and construction of superconducting magnets has evolved dramatically over the last three decades leading to an overall increase in their complexity. Their use has moved from particle physics through medical applications such as MRI to the very large magnets demanded by fusion technology, where the optimal quality of the HV insulation is a critical step in achieving the design performance. The Vacuum Pressure Impregnation (VPI) technique-a particularly refined case of the more widely known Vacuum Assisted Resin Transfer Mould (VARTM)-has become the most common process for the consolidating electrical insulation of large superconducting magnets. To achieve success with the VPI process demands that detailed attention is paid to many steps, namely: coil drying, resin degassing and coil impregnation. Commercial sensitivity hasmeant that there has not been a wellpublished exchange of information and therefore many details of the VPI process may vary in different organizations. The present paper aims at filling this gap and will discuss in detail: 1) all required steps with the main risks in each of them, 2) commonly used methods in each of the steps for optimized control of the process and 3) tooling to minimize risk of failure during impregnation. © 2011 IEEE.
De Tommasi G.,University of Naples Federico II |
Neto A.C.,Fusion for Energy F4E |
Sterle C.,University of Naples Federico II
IEEE Transactions on Plasma Science | Year: 2014
The magnetic diagnostic system plays a crucial role in a Tokamak reactor and, more generally, in any magnetically confined fusion device. Indeed, all the essential quantities for plasma control are computed by a number of processing nodes, starting from the magnetic fluxes and fields acquired by this diagnostic system. This paper presents the tool for oPtImal Measurement Probes Allocation (PIMPA) in a magnetic diagnostic system. PIMPA is based on the solution of an integer linear programming problem, and aims at maximizing the reliability of the diagnostic system against the failure of one or more of the processing nodes. Although in this paper, PIMPA is introduced for the optimal allocation among the processing nodes of the magnetic probes, it can be easily extended to any diagnostic system. A case study that considers a distributed and scalable architecture for the ITER Tokamak is presented to show the effectiveness of the proposed approach. © 2014 IEEE.
Rivas J.-C.,Fusion for Energy F4E |
Dies J.,Fusion for Energy F4E
Fusion Science and Technology | Year: 2011
In this contribution, an upgraded model for plasma-wall interaction in the AINA code is presented. The AINA code is a comprehensive hybrid code comprising a global balance plasma dynamics model and a radial and poloidal thermal analysis of in-vessel components. AINA is an evolution of the SAFALY code, which was initially adopted to assess ITER EDA plasma safety events and quantitatively investigate plasma instability events in nuclear fusion reactors such as ITER. The new erosion code module includes algorithms for the most relevant plasma wall interaction phenomena that will take place in the ITER vessel during the steady state of the normal operation. Physical sputtering, radiation enhanced sublimation (RES), and chemical erosion algorithms have been added to the previous thermal sublimation algorithm. The erosion results from these models have been benchmarked with results for ITER normal operation from the B2-Eirene code. The new erosion model had to be tested with external data for particle fluxes over the wall, because the AINA code does not presently have the ability to model those particle fluxes. However, with the new results, the impurity transport model parameters have been re-calibrated and some useful conclusions have been extracted.
Bindslev H.,Fusion for Energy F4E
Fusion Engineering and Design | Year: 2015
Fusion for Energy (F4E), on behalf of Europe, is responsible for the procurement of most of the high-technology items for the ITER device. This paper provides an overview of the technical status of Europe's contributions to ITER and the related challenges. In particular, we report on progress in the construction of the buildings at the Cadarache site, the fabrication of the superconducting magnets and the vacuum vessel and the testing and qualification of the in-vessel components (first wall and divertor). The status of the design and development of the additional heating systems and the test blanket modules will also be described. © 2015 Elsevier B.V. All rights reserved.
Portone A.,Fusion for Energy F4E
Proceedings of the IEEE Conference on Decision and Control | Year: 2012
In this paper a typical optimal control problem arising in tokamak engineering is presented and solved. The attention is focused on the selection of the optimal number and positions of the Poloidal Field (PF) coils to keep in MHD equilibrium a tokamak plasma of given characteristics. The problem is casted as a sparse optimal least square problem and it is solved by using a QR factorizations with pivoting. A simple case study is applied to a tokamak configuration of the ITER-class type. © 2012 IEEE.
Rivas J.C.,Fusion for Energy F4E |
Dies J.,Fusion for Energy F4E
Fusion Engineering and Design | Year: 2013
The loss of plasma control events in ITER are safety cases investigated to give an upper bound of the worse effects foreseeable from a total failure of the plasma control function. In the past, conservative analyses based on simple 0D models for plasma balance equations and 1D models for wall heat transfer showed that a hypothetical scenario of first wall coolant tubes melting and subsequent entering of water in the vacuum vessel could not be totally excluded. AINA (Analyses of IN vessel Accidents) code is a safety code developed at Fusion Energy Engineering Laboratory (FEEL) in Barcelona. It uses a 0D-1D architecture, similar to that used for previous analyses of ITER loss of plasma control events. The results of this study show the simultaneous effect of two perturbations (overfuelling and overheating) over a plasma transient, and compare it with the isolated effects of each perturbation. It is shown that the combined effect can be more severe, and a method is outlined to locate the most dangerous transients over a nT diagram. © 2013 Elsevier B.V.
Ingesson L.C.,Fusion for Energy F4E
AIP Conference Proceedings | Year: 2014
A project-oriented approach is beneficial for the selection and design of viable diagnostics for fusion reactors because of the associated complex physical and organizational environment. The project-oriented approach includes rigorous risk management. The nature and impact of risks related to technical, organizational and commercial aspects in relation to the development of ITER diagnostics under EU responsibility are analyzed. The majority of risks are related to organizational aspects and technical feasibility issues. The experience with ITER is extrapolated to DEMO and beyond. It should not be taken for granted that technical solutions will be found, while a risk analysis of various diagnostic techniques with quantitative assessments undertaken early in the design of DEMO would be beneficial. © 2014 American Institute of Physics.
Portone A.,Fusion for Energy F4E
IEEE Transactions on Applied Superconductivity | Year: 2014
The aim of this paper is to present an optimization procedure for the design of the Poloidal Field (PF) coil system of a tokamak reactor. Starting from the target plasma shape, a simplified toroidal field (TF) coil shape is derived to satisfy the basic requirements of bending-free coil and maximum (usually about 1%) TF ripple at the plasma boundary. Then, the PF coils, which are mechanically attached to the TF, optimal number, and locations, are computed. This computation is done by optimally matching at the desired plasma boundary the external magnetic flux that keeps the plasma magneto hydro dynamic (MHD) equilibrium with the magnetic flux generated by the PF coils. The external magnetic flux is matched at the plasma boundary by optimally selecting the principal components of the modified mutual inductance matrix M∗ that minimize the plasma boundary deformation from its reference shape. Since the M∗ accounts for the plasma response through the linearization of the Grad-Shafranov equation, the solution found is correct to first order and it substantially simplifies the lengthy, nonlinear computations that are presently used to achieve the same goal. © 2002-2011 IEEE.