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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. Source


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. Source


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. Source


Franza F.,Polytechnic University of Turin | Ciampichetti A.,ENEA | Ricapito I.,Fusion for Energy F4E | Zucchetti M.,Polytechnic University of Turin
Fusion Engineering and Design | Year: 2012

Hydrogen dissolves in and permeates through most materials, thus it is important to understand the permeation, diffusion and dissolution phenomena of atomic hydrogen in materials in which hydrogen and its isotopes are present. In this work the problem of tritium transport from lead-lithium breeder through different heat transfer surfaces to the environment has been studied and analyzed by means of a computational code. The code (FUS-TPC) is a new fusion-devoted version of the fast-fission one called Sodium-Cooled Fast Reactor Tritium Permeation Code (SFR-TPC). The main features of the model inside the code are described. A simulation, using the code, was performed by adopting the configuration of the European configuration of the Helium Cooled Lead Lithium (HCLL) blanket for DEMO. © 2012 Elsevier B.V. All rights reserved. Source


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. Source

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