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Merate, Italy

Di Gironimo G.,University of Naples Federico II | Labate C.,University of Naples Federico II | Renno F.,University of Naples Federico II | Brolatti G.,ENEA | And 5 more authors.
Fusion Engineering and Design

The paper presents a concept design of a remote handling (RH) system oriented to maintenance operations on the divertor second cassette in FAST, a satellite of ITER tokamak. Starting from ITER configuration, a suitably scaled system, composed by a cassette multifunctional mover (CMM) connected to a second cassette end-effector (SCEE), can represent a very efficient solution for FAST machine. The presence of a further system able to open the divertor port, used for RH aims, and remove the first cassette, already aligned with the radial direction of the port, is presumed. Although an ITER-like system maintains essentially shape and proportions of its reference configuration, an appropriate arrangement with FAST environment is needed, taking into account new requirements due to different dimensions, weights and geometries. The use of virtual prototyping and the possibility to involve a great number of persons, not only mechanical designers but also physicist, plasma experts and personnel assigned to remote handling operations, made them to share the multiphysics design experience, according to a concurrent engineering approach. Nevertheless, according to the main objective of any satellite tokamak, such an approach benefits the study of enhancements to ITER RH system and the exploration of alternative solutions. © 2013 Elsevier B.V. Source

Crescenzi F.,ENEA | Roccella S.,ENEA | Brolatti G.,ENEA | Cao L.,CAS Hefei Institutes of Physical Science | And 8 more authors.
Fusion Engineering and Design

The Fusion Advanced Study Torus (FAST), with its compact Tokamak design, high toroidal field and plasma current, will face many of the problems that ITER will meet and will anticipate many DEMO relevant physics and technology issues. The Design Upgrade of the Vessel and In-Vessel Components is presented in this paper. Relevant modifications were performed to the Vacuum Vessel (VV) and to the Plasma Facing Components (PFCs), i.e. the First Wall (FW) and the Divertor. The VV was modified to insert active reduction coils (ARC), between VV and the toroidal field (TF) coils to keep toroidal field magnet ripple lower than 0.3% and to allow Remote Handling for the FW and the Divertor. The FW, was modified to house coils for ELMs control and other plasma instabilities. A 3D thermo-hydraulic analysis using ANSYS code was performed to check FW heat removal capability. A new Divertor was designed to withstand the largest thermal loads of the high performance, low density, H-mode and to be able to comply with a recent magnetic topology called as "Snow Flake", increasing up a factor 4 the flux expansion. An exhaustive 3D thermo-hydraulic analysis using ANSYS code was carried out to show the capability of the Divertor to comply these high requirements. Design criteria were satisfied by present components of the upgraded machine. © 2013 Elsevier B.V. Source

Grisham L.R.,Princeton Plasma Physics Laboratory | Agostinetti P.,Consorzio RFX | Barrera G.,CIEMAT | Blatchford P.,Culham Center for Fusion Energy | And 29 more authors.
Fusion Engineering and Design

The ITER [1] fusion device is expected to demonstrate the feasibility of magnetically confined deuterium-tritium plasma as an energy source which might one day lead to practical power plants. Injection of energetic beams of neutral atoms (up to 1 MeV D 0 or up to 870 keV H 0) will be one of the primary methods used for heating the plasma, and for driving toroidal electrical current within it, the latter being essential in producing the required magnetic confinement field configuration. The design calls for each beamline to inject up to 16.5 MW of power through the duct into the tokamak, with an initial complement of two beamlines injecting parallel to the direction of the current arising from the tokamak transformer effect, and with the possibility of eventually adding a third beamline, also in the co-current direction. The general design of the beamlines has taken shape over the past 17 years [2], and is now predicated upon an RF-driven negative ion source based upon the line of sources developed by the Institute for Plasma Physics (IPP) at Garching during recent decades [3-5], and a multiple-aperture multiple-grid electrostatic accelerator derived from negative ion accelerators developed by the Japan Atomic Energy Agency (JAEA) across a similar span of time [6-8]. During the past years, the basic concept of the beam system has been further refined and developed, and assessment of suitable fabrication techniques has begun. While many design details which will be important to the installation and implementation of the ITER beams have been worked out during this time, this paper focuses upon those changes to the overall design concept which might be of general interest within the technical community. Source

Testoni P.,Fusion for Energy F4E | Albanese R.,ENEA | Lucca F.,LT Calcoli SaS | Roccella M.,LT Calcoli SaS | And 4 more authors.
Fusion Engineering and Design

This paper presents the results of the electromagnetic analyses of the ITER vacuum vessel and blanket modules. A wide collection of electromagnetic transients has been simulated: VDEs and MDs belonging to cat III, II and I, and a magnet fast discharge. Eddy and halo currents and corresponding Lorentz forces have been computed using 3D solid FE models implemented in ANSYS and CARIDDI. The plasma equilibrium configurations (displacement and quench of the plasma current, toroidal flux variation due to the β drop and halo currents wetting the first wall) used as an input for the EM analyses have been supplied by the 2D axisymmetric code DINA. The paper describes in detail the methodology used for the analyses and the main results obtained. © 2012 Elsevier B.V. Source

Encheva A.,ITER Organization | Omran H.,United Technologies | Perez-Lasala M.,Fusion for Energy F4E | Alekseev A.,D.V. Efremov Scientific Research Institute of Eelectrophysical Apparatus | And 26 more authors.
Fusion Engineering and Design

Diagnostic electrical services provide the electrical infrastructure to serve diagnostic components installed on the ITER tokamak. This infrastructure is composed of cables, connectors, cable tails, looms, conduits and feedthroughs. The diagnostic services offer as well a shelter for various instrumentations - vacuum vessel (VV), blanket and divertor. The diagnostic sensors are located on the inner and outer VV wall, on blanket shield modules, divertor cassettes and in port plugs. They require electrical cabling to extract the measurement and, in some cases, to supply electrical power to the sensors. These cables run from the sensors to feedthroughs on the VV and the port interspace or cryostat. The design and integration of all components that are part of diagnostic electrical services is an important engineering activity that is being challenged by the multiple requirements and constraints which have to be satisfied while at the same time delivering the required diagnostic performance. The positioning of the components must correlate not only with their functional specifications but also with the design of the major ITER components. This is a particular challenge because not all systems have reached the same level of design maturity. This paper outlines the engineering challenges of ITER diagnostics electrical services. The environmental conditions inside the VV will have an important impact. Leading risks to these components include poor electrical contact at connectors, the effects of exposure to nuclear irradiation, such as material transmutation, heating, and generation of spurious electrical signals etc., failure due to electromagnetic forces and electrical interference due to the noisy environment. Last but not least are the challenges for confinement and vacuum requirements set up on electrical feedthroughs. It will focus as well on the design and structural assessment of all components, and their requirements. Besides the integration limitations, the loads are the main design driver. © 2013 Elsevier B.V. All rights reserved. Source

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