Consorzio CREATE

Napoli, Italy

Consorzio CREATE

Napoli, Italy

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Wenninger R.,EUROfusion Programme Management Unit | Wenninger R.,Max Planck Institute for Plasma Physics (Garching) | Albanese R.,University of Naples Federico II | Ambrosino R.,Parthenope University of Naples | And 36 more authors.
Nuclear Fusion | Year: 2017

For several reasons the challenge to keep the loads to the first wall within engineering limits is substantially higher in DEMO compared to ITER. Therefore the pre-conceptual design development for DEMO that is currently ongoing in Europe needs to be based on load estimates that are derived employing the most recent plasma edge physics knowledge. An initial assessment of the static wall heat load limit in DEMO infers that the steady state peak heat flux limit on the majority of the DEMO first wall should not be assumed to be higher than 1.0 MW m-2. This compares to an average wall heat load of 0.29 MW m-2 for the design assuming a perfect homogeneous distribution. The main part of this publication concentrates on the development of first DEMO estimates for charged particle, radiation, fast particle (all static) and disruption heat loads. Employing an initial engineering wall design with clear optimization potential in combination with parameters for the flat-top phase (x-point configuration), loads up to 7 MW m-2 (penalty factor for tolerances etc not applied) have been calculated. Assuming a fraction of power radiated from the x-point region between 1/5 and 1/3, peaks of the total power flux density due to radiation of 0.6-0.8 MW m-2 are found in the outer baffle region. This first review of wall loads, and the associated limits in DEMO clearly underlines a significant challenge that necessitates substantial engineering efforts as well as a considerable consolidation of the associated physics basis. © 2017 Euratom.


Crescenzi F.,ENEA | Marini F.,ENEA | Nardi C.,ENEA | Pizzuto A.,ENEA | And 5 more authors.
Fusion Engineering and Design | Year: 2011

Pre-compressions rings have been designed to improve the ITER magnets structural support. They are made of high strength unidirectional S-2 glass fibres wound in an epoxy matrix. In order to obtain very high strength properties, ENEA developed and characterized this composite with a high volumetric glass content (∼68%). Benefits of this solution include high strength, no interference with magnet fields and long expected service life. This work illustrates the mechanical characterization of the material used to manufacture the rings. At first, linear specimens were produced to perform tensile and creep tests. Then, reduced scale ring mock-ups have been fabricated and tested in ENEA Fusion-Laboratories. One ring was used for the machining of several specimens for compression and shear tests. Standard and non standard specimens were machined from different geometric directions (longitudinal, radial and transversal with reference to the fibre-glass direction) and tested at room temperature and at liquid nitrogen (77 K) temperature to complete characterization with compression and shear tests. The experimental campaign has been carried out following as close as possible the related ASTM standards in order to evaluate material strength, Young and shear moduli. Test results showed high mechanical strength of the composite in fibre-glass longitudinal direction but lower values in transversal direction. © 2011 EURATOM ENEA Association-ENEA Fusion Unit.


Rossi P.,ENEA | Capobianchi M.,ENEA | Crescenzi F.,ENEA | Massimi A.,ENEA | And 7 more authors.
Fusion Engineering and Design | Year: 2011

ENEA has developed and characterized a high strength glass fibre-epoxy composite as reference material for the manufacture of the two sets of 3 pre-compression rings located at top and bottom of the inner straight leg region of the ITER Toroidal Field (TF) coils. These rings will provide a radial force of about 70 MN/coil at cryogenic temperature pulling the TF coils into contact and reducing toroidal tension in the four outer intercoil structures. The paper describes the ultimate tensile strength (UTS) testing campaign carried out at ENEA Frascati laboratories on six different rings manufactured winding S2 glass fibers on a diameter of 1 m (1/5 of the full scale) by both vacuum pressure epoxy impregnation and filament wet winding techniques. The volumetric glass content was around 70%. The rings were expanded with radial steps of 0.1 mm into a dedicated hydraulic testing machine consisting of 18 radial actuators working in position control with a total capability of 1000 tons. All the mock-ups showed very high tensile strength (1550 MPa is the average of the mean hoop stresses at failure) and a practically constant tensile modulus. The test results are reported and discussed. © 2011 EURATOM ENEA Association-ENEA Fusion Unit. Published by Elsevier B.V. All rights reserved.


Knaster J.,ITER Organization | Amoskov V.,D.V. Efremov Scientific Research Institute of Eelectrophysical Apparatus | Formisano A.,Consorzio CREATE | Gribov Y.,ITER Organization | And 7 more authors.
Fusion Engineering and Design | Year: 2011

Error fields in Tokamaks are small departures of the exact axisymmetry of the ideal magnetic field configuration. Their reduction beyond a threshold value by the error field correction coils is essential since sufficiently large static error fields lead to discharge disruption. The error fields are originated not only by coils fabrication and installation alignment tolerances, joints and busbars but also due to the presence of ferromagnetic elements. The start of plasma current flattop with relatively low plasma density is considered as a critical state of the 15 MA scenario for the onset of locked modes causing disruptions. A figure of merit, the '3-mode' criterion, based on the lowest error field harmonics [(1,1); (2,1); (3,1)] was chosen to assess the error fields expected in ITER. Analysis performed last years in an independent way by CREATE (EU) and Efremov Inst. (RF) groups has allowed a deep understanding of the error fields induced by all the possible sources. Both groups were successfully benchmarked with the estimation of the error fields induced by a given deformed shape of a TF coil obtaining a perfect match of the results up to 2 orders of magnitude smaller than the defined threshold of B 3-mode/Bto < 5 × 10-5. Three different sets of independent variables based on a 3D rigid body movement of the coils have been provided for the 6 CS modules, 18 TF coils and 6 PF coils tolerances, which have allowed a clear understanding of the weight of each of the variables in the induced error fields. The results obtained in 2008 with a realistic set of magnets tolerances concluded that the system of correction coils provides effective suppression of the error fields with margin to correct possible impact of other sources. In particular, it has been shown that superconducting joints, feeders and busbars play a secondary effect; however the radial position of the TF coils and the tilt and radial shift of the CS stack would have a relevant influence. The ensuing recent sets of variables studied aimed at deepening the understanding of the operational limits. The present paper addresses the impact of the fabrication tolerances and installation misalignments of the magnet system; the impact of ferromagnetic inserts and busbars as well as the design of the correction coils will be thoroughly covered elsewhere (Amoskov, et al., in press [1]). © 2011 Elsevier B.V. All Rights Reserved.


Testoni P.,Fusion for Energy F4E | Oliva A.B.,Fusion for Energy F4E | Portone A.,Fusion for Energy F4E | Carin Y.,Fusion for Energy F4E | And 8 more authors.
Fusion Engineering and Design | Year: 2011

Fusion for energy (F4E), the European Domestic Agency for ITER, is involved in a relevant number of activities in the area of electromagnetic analysis in support of ITER general design, and of specific requirement of the EU in-kind procurement. In this context, its main activity is linked with the electromagnetic analysis of several ITER components (blanket shield modules and first wall panels, blanket cooling manifolds, TBM port plug, etc.) subjected to electro-dynamical loads. Another important activity is related to the ITER superconducting magnets, namely the quench detection of the ITER TF coils and the Joule losses in the magnets cold structures. Last, but not least, a further activity is on going on the error field analysis due to tolerances in both construction and assembly of ITER magnets. © 2011 Elsevier B.V. All rights reserved.


Cau F.,Fusion for Energy F4E | Bessette D.,ITER Organization | D'Amico G.,Fusion for Energy F4E | Portone A.,Fusion for Energy F4E | And 4 more authors.
IEEE Transactions on Applied Superconductivity | Year: 2016

The ITER magnet system, which is composed of toroidal field (TF) and poloidal field (PF) coils, central solenoid (CS), correction coils, and all their structural supports, is cooled by supercritical helium flow at 4.5 K. Eddy currents are induced in the metallic components during normal operation, due to the variation of PF and CS coils current, and to the plasma ramp-up and ramp-down. In addition, during disruptions, as well as other fast plasma transient events, the eddy currents circulating in these structures reach very high values due to the high induced electric fields. An effective cooling is therefore needed to limit the temperature increase of the magnets and their supports. The Joule energy dissipated in the cold structures of the ITER magnet system has been computed by means of the electromagnetic finite-element code CARIDDI. A model of a 40 degree sector of the ITER magnet has been built in order to represent in detail the connections between the CS and the TF coils, PF and CS supports, and all the intercoil structures and the metallic part of the TF coils (radial plates, case, etc.). Vacuum vessel, thermal shield, and cryostat have also been modeled. The reference 15-MA inductive scenario (with plasma current ramp-up duration of 80 s and ramp-down of 200 s) and a 15-MA fast inductive scenario characterized by the fastest plasma current ramp-up (50 s) and the fastest plasma current ramp-down (65 s) have been analyzed. In addition, several plasma instabilities have been simulated considering the reference electrical connections between the metallic parts. © 2016 IEEE.


Ravera G.L.,ENEA | Ceccuzzi S.,Third University of Rome | Cardinali A.,ENEA | Cesario R.,ENEA | And 3 more authors.
AIP Conference Proceedings | Year: 2014

The preliminary assessment of a Lower Hybrid Current Drive (LHCD) system for the DEMOnstration power plant (DEMO) is mainly focused on the R&D needs of the less conventional RF components of the Main Transmission Line (MTL) and of the launcher. 500 kW, CW klystrons will be used to deliver the RF power to independent Passive Active Multijunction (PAM) launcher modules at 5 GHz. This paper describes the criteria followed to investigate the optimum solution for the RF window used as vacuum barrier between the MTL and the launcher, an open issue in the LHCD system for ITER too. The best candidate, capable of withstanding a power level of, or above, 0.5 MW in CW operation and to satisfy the electrical and thermonuclear requirements, is a Pill-Box assembly, based on a thin single disk of CVD-diamond as dielectric, water cooled at the edge. A thickness of 3 mm, much shorter than half a wavelength of the TE°11 mode in the dielectric as in the conventional window (unfeasible and too expensive with CVD-diamond at these frequencies), is sufficient to limit the exerted stress at the edge under the fracture stress for a maximum pressure applied of 0.9 MPa. In this paper the simulation results of conventional and thin CVD-diamond vacuum windows are presented comparing S-parameters, losses and electric fields in both matching condition and with VSWR = 2, using WR284 and WR229 as input/output rectangular waveguide. © 2014 American Institute of Physics.


Mirizzi F.,Consorzio CREATE | Spassovsky I.,ENEA | Ceccuzzi S.,ENEA | Dattoli G.,ENEA | And 9 more authors.
Fusion Engineering and Design | Year: 2015

ECRH&CD systems are extensively used on tokamak plasmas due to their capability of highly tailored power deposition, allowing very localised heating and non-inductive current drive, useful for MHD and profiles control. The high electron temperatures expected in DEMO will require ECRH systems with operating frequency in the 200-300 GHz range, equipped with a reasonable number of high power (P ≥ 1 MW) CW RF sources, for allowing central RF power deposition. In this frame the ENEA Fusion Department (Frascati) is coordinating a task force aimed at the study and realisation of a suitable high power, high frequency reliable source. © 2014 Elsevier B.V. All rights reserved.


Ravera G.L.,ENEA | Ceccuzzi S.,ENEA | Doria A.,ENEA | Spassovsky I.,ENEA | And 2 more authors.
European Microwave Week 2015: "Freedom Through Microwaves", EuMW 2015 - Conference Proceedings; 2015 45th European Microwave Conference Proceedings, EuMC | Year: 2015

Two configurations of mode converters from the TE10 in rectangular waveguide to the TE82 in circular waveguide are designed. Such conversion is required to perform the cold test of a Bragg resonator for a 250 GHz Cyclotron Auto-Resonance Maser under development at the ENEA research center. The cylindrical cavity to be tested is far more oversized than any previous CARM experiments, making the design of mode converters particularly challenging. One conversion chain uses an in-line coupling beat-wave TE11/TE81 converter, whereas the other is a sidewall coupling mode converter. Both require a final, highly oversized, step-Type converter from the TE81 to the TE82 mode; its performances are computed with both a commercial software and an in-house mode matching code, whose results agree fairly well. For both configurations, simulation results are presented and the most promising solution is identified. © 2015 EuMA.


Cucchiaro A.,ENEA | Albanese R.,Consorzio CREATE | Ambrosino G.,Consorzio CREATE | Brolatti G.,ENEA | And 14 more authors.
Fusion Engineering and Design | Year: 2010

Fusion advanced studies torus (FAST) is a proposal for a satellite facility which can contribute the rapid exploitation of ITER and prepare ITER and DEMO regimes of operation, as well as exploiting innovative DEMO technology. FAST is a compact (R0 = 1.82 m, a = 0.64 m, triangularity δ = 0.4) machine able to investigate non-linear dynamics effects of alpha particle behaviours in burning plasmas [1,2,5]. The project is based on a dominant 30 MW of ion cyclotron resonance heating (ICRH), 6 MW of lower hybrid (LH) and 4 MW of electron cyclotron resonance heating (ECRH). FAST operates at a wide range [3,4] of parameters, e.g., in high performance H-mode (BT up to 8.5 T; IP up to 8 MA) as well as in advanced Tokamak operation (IP = 3 MA), and full non-inductive current scenario (IP = 2 MA). Helium gas at 30 K is used for cooling the resistive copper magnets [6]. That allows for a pulse duration up to 170 s. To limit the TF magnet ripple ferromagnetic insert have been introduced inside the vacuum vessel (VV). Ports have been designed to also accommodate up to 10 MW of negative neutral beam injection (NNBI). Tungsten (W) or liquid lithium (L-Li) have been chosen as the divertor plates material, and argon or neon as the injected impurities to mitigate the thermal loads. © 2009 Elsevier B.V. All rights reserved.

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