A. Torre

CEA Cadarache

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Papers

Citation

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i-10 index

Research Trends

Author Order

Document Type

Co-Authors

L. Zani

CEA Cadarache

Benoît Lacroix

CEA Cadarache

S. Nicollet

CEA Cadarache

D. Ciazynski

Commissariat à l'Énergie Atomique et aux Énergies Alternatives

P. Decool

Commissariat à l'Énergie Atomique et aux Énergies Alternatives

Quentin Le Coz

CEA Cadarache

F. Nunio

Commissariat à l'Énergie Atomique et aux Énergies Alternatives

R. Vallcorba

Institut de Recherche sur les Lois Fondamentales de l'Univers

A. Louzguiti

ITER

G. Jiolat

CEA Cadarache

Cooperative Institutions

Commissariat à l'Énergie Atomique et aux Énergies Alternatives

769

CEA Cadarache

767

Institut de Recherche sur la Fusion par Confinement Magnétique

316

Centre National de la Recherche Scientifique

186

Culham Science Centre

178

Max Planck Institute for Plasma Physics

139

ITER

126

Royal Military Academy

119

National Agency for New Technologies, Energy and Sustainable Economic Development

111

Aix-Marseille Université

101

Author Statistics

Papers

Citation

H-Index

i-10 index

Research Field

Coupled Mechanical and Electrical Modeling of Nb₃Sn Strand Critical Current Under Bending

IEEE Transactions on Applied Superconductivity (2016)

D. Ciazynski A. Torre S. Li Gilles Lenoir

Strain dependence of Nb ₃ Sn superconducting properties is known to be responsible for the degradation of transport current capability of large steel jacketed cable-in-conduit conductors (CICCs). The mechanical deformations of the strands in the cables due to both cooldown after heat treatment and Lorentz force during operation are the main sources of strand-in-cable critical current degradation. The complete modeling of a CICC relies first on the modeling of the single strand with its superconducting filaments and then on the modeling of the strands in the cable. This paper reports on the description of the models and the analyses performed using coupled mechanical and electrical modeling. The key of this modeling lies in its ability to characterize the superconducting properties of the twisted Nb ₃ Sn filaments inside a locally bent strand from a global mechanical deformation map, taking into account possible current redistribution among filaments all along the strand length. In the same way, strands may experience local critical current degradation in a CICC, and current transfer between strands at their contact points must be modeled to predict cable performance. This work can be seen as the first stones of an integrated coupled mechanical-electrical modeling of large CICCs made with strain-sensitive strands.

Niobium-tin

10.1109/tasc.2016.2535152

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Citations (3)

Performance of JT-60SA toroidal field coils in light of strand and conductor test results

Fusion Engineering and Design (2017)

D. Ciazynski A. Torre L. Zani P. Decool S. Nicollet

Toroidal field

10.1016/j.fusengdes.2016.12.028

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Citations (9)

Science and technology research and development in support to ITER and the Broader Approach at CEA

Nuclear Fusion (2013)

A. Bécoulet G.T. Hoang J. Abiteboul J. Achard Tomás Alarcón

In parallel to the direct contribution to the procurement phase of ITER and Broader Approach, CEA has initiated research & development programmes, accompanied by experiments together with a significant modelling effort, aimed at ensuring robust operation, plasma performance, as well as mitigating the risks of the procurement phase. This overview reports the latest progress in both fusion science and technology including many areas, namely the mitigation of superconducting magnet quenches, disruption-generated runaway electrons, edge-localized modes (ELMs), the development of imaging surveillance, and heating and current drive systems for steady-state operation. The WEST (W Environment for Steady-state Tokamaks) project, turning Tore Supra into an actively cooled W-divertor platform open to the ITER partners and industries, is presented.

Tore Supra

Bootstrap current

Limiter

10.1088/0029-5515/53/10/104023

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Citations (8)

WEST full tungsten operation with an ITER grade divertor

Nuclear Fusion (2024)

J. Bucalossi A. Ekedahl J. Achard K. Afonin O. Agullo

Abstract The mission of WEST (tungsten-W Environment in Steady-state Tokamak) is to explore long pulse operation in a full tungsten (W) environment for preparing next-step fusion devices (ITER and DEMO) with a focus on testing the ITER actively cooled W divertor in tokamak conditions. Following the successful completion of phase 1 (2016-2021), phase 2 started in December 2022 with the lower divertor made entirely of actively cooled ITER-grade tungsten mono-blocks. A boronization prior the first plasma attempt allowed for a smooth startup with the new divertor. Despite the reduced operating window due to tungsten, rapid progress has been made in long pulse operation, resulting in discharges with a pulse length of 100 s and an injected energy of around 300 MJ per discharge. Plasma startup studies were carried out with equatorial boron nitride limiters to compare them with tungsten limiters, while Ion Cyclotron Resonance Heating assisted startup was attempted. High fluence operation in attached regime, which was the main thrust of the first campaigns, already showed the progressive build up of deposits and appearance of dust, impacting the plasma operation as the plasma fluence increased. In total, the cumulated injected energy during the first campaigns reached 43 GJ and the cumulated plasma time exceeded 5 h. Demonstration of controlled X-Point Radiator regime is also reported, opening a promising route for investigating plasma exhaust and plasma-wall interaction issues in more detached regime. This paper summarises the lessons learned from the manufacturing and the first operation of the ITER-grade divertor, describing the progress achieved in optimising operation in a full W environment with a focus on long pulse operation and plasma wall interaction.

Limiter

Fusion power

Thermonuclear Fusion

10.1088/1741-4326/ad64e5

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Citations (2)

Updates on Magnet Design For EU-DEMO Reactor: Optimization Studies on TF and CS Systems

IEEE Transactions on Applied Superconductivity (2021)

L. Zani François Bonne V. Corato Q. Le Coz B Aguiar Jose

In the framework of EU design activities for dimensioning the future fusion DEMOnstration reactor (DEMO), extensive analyses were conducted in EUROfusion context, aiming at ultimately defining the design of the DEMO magnets system. In this objective CEA (Commissariat à l'Énergie Atomique et aux Énergies Alternatives) proposes design for both Central Solenoid (CS) and Toroidal Field (TF) cryomagnetic systems. Considering the latest DEMO reactor baseline both systems were investigated using pre-dimensioning CEA tool MADMACS, with an outcome of several options for each of them. In a second step those designs were investigated with detailed analyses (electromagnetic, thermohydraulic, mechanic). In the paper a broad analysis of the CS design considering fatigue constraints is shown, together with recommendations considering flux target. The TF system considered is an ITER-like option with radial plates and the detailed electromagnetic analysis is exposed and discussed in comparison with former methods. The outcome is used for a thermohydraulic detailed static analysis and a mechanical analysis of TF winding pack is provided and discussed in terms of results but of method as well. Global considerations and recommendations will also be provided to give a first view on cryomagnetic system optimization with respect to cost merits related to both magnet systems.

Solenoid

Dimensioning

Fusion power

Toroidal field

10.1109/tasc.2021.3068560

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Citations (6)

OLYMPE, a multi-physic platform for fusion magnet design: Development status and first applications

Cryogenics (2020)

L. Zani Benoît Lacroix A. Torre Jean-Claude Vallet C. Berthinier

Dimensioning

10.1016/j.cryogenics.2020.103086

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Citations (9)

Study of the ITER TF CICC Mechanical Behavior Under Cool-Down and Repetitive EM Loadings

IEEE Transactions on Applied Superconductivity (2021)

Rebecca Riccioli A. Torre Damien Durville Marco Breschi Frédéric Lebon

High-current and high-field superconducting magnets sustain huge electromagnetic loads. On top of that, the high field constraint pushes towards the use of high-performing superconducting Nb ₃ Sn strands, which are strain sensitive. In recent fusion magnets, the loads reach few hundreds of kN/m in static and sometimes cyclic regimes. These loads modify the strands local strain state and the overall mechanical properties of the cable. In this article, the finite element mechanical code MULTIFIL is applied to model an ITER Cable-In-Conduit Conductor (CICC) petal for the Toroidal Field (TF) coil under operating loadings. The local strain maps and the macroscopic mechanical behavior of the cable are analyzed for both thermal and electromagnetic loadings. A possible explanation for the initial drop of the current sharing temperature observed in some of the TF CICCs tested in the SULTAN facility (PSI, Villigen) is proposed. The impact of important model parameters, such as the void fraction and the friction coefficient between strands, on the mechanical behavior of the conductor is investigated.

Fusion power

10.1109/tasc.2021.3070988

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Citations (2)

CVN impact energy and fracture characteristics correlations with different oxide nanoparticles improving submerged arc welds

Materials Research Express (2021)

A. Jiménez Ana María Paniagua-Mercado Arturo García‐Bórquez A. Torre C. Mejía‐García

Abstract The main aim of this research is to correlate the toughness improvement with fracture features of tested Charpy V-Notch impact specimens obtained from submerged arc welds enriched by different oxide nanoparticles (TiO 2 , SiO 2 , Al 2 O 3 and Mn 2 O 3 ). Through Scanning Electron Microscopy (SEM) fractography, fracture features were determined, which are the shear fracture percent, the loading angle, as well as the diameter of dimples and the width of cleavage sheets at ductile and brittle fracture regions, respectively. It was observed an increase of the Charpy V-Notch impact energy with the decrease of loading angle and the average diameter of dimples, as well as with the increase of its shear fractures percents. Moreover, it was correlated the increase of Acicular Ferrite surface density with the decrease of width of cleavage sheets. Thus, the toughness of the submerged arc welds can be improved by adding TiO 2 , SiO 2 , Al 2 O 3 or Mn 2 O 3 nanoparticles.

Charpy impact test

Fractography

Acicular ferrite

Dimple

Brittleness

Cleavage (geology)

10.1088/2053-1591/abdaf2

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Citations (5)

Preparation of PF1/6 and PF2 Conductor Performance Qualification Sample

IEEE Transactions on Applied Superconductivity (2010)

L. Reccia S. Turtù Gian Mario Polli L. Affinito F. Maierna

The ITER Conductor Procurement Arrangement requires that all the conductors have to be qualified before the production phase. The objects of the work here presented are the Poloidal Field Coils conductors, specifically the PF1/6 and the PF2 conductors. The realization of the samples comprised the jacketing, the compaction and the straightening of the CICC, together with the design and manufacturing of the bottom hairpin box and the upper termination. The instrumentation and the related jacket machining completed the preparation of the samples. The samples have been assembled according to the specification defined by ITER and have been shipped to SULTAN facility. This paper describes all the activities performed during the preparation, the features of the samples and all the related issues.

Instrumentation

Realization (probability)

10.1109/tasc.2010.2081963

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Citations (8)

Experimental and Analytical Approaches on JT—60SA TF Strand and TF Conductor Quality Control During Qualification and Production Manufacture Stages

IEEE Transactions on Applied Superconductivity (2012)

L. Zani P. Barabaschi P. Bruzzone H. Cloez E. Di Pietro

In the framework of the JT-60SA project, aiming at upgrading the present JT-60U tokamak, Europe as part of its in-kind contribution within the Broader Approach agreement, will provide the full Toroidal Field (TF) magnet system. For this purpose, Fusion for Energy is committed to procure about 27 km of TF conductor. The TF conductor is cable-in-conduit type and includes 486 strands (2/3 NbTi-1/3 copper) wrapped with a thin stainless steel foil and embedded into a rectangular stainless steel jacket. The procurement is split into two main contracts: one for strand manufacturing and the other for cabling and jacketing. After having successfully passed the qualification stage, strand and conductor are now in mass production stage. In the present paper, we emphasize on the scientific and quality control approaches of both TF strand and TF conductor productions. For the NbTi strand, the focus is on design and measurement of performance critical aspects (CuNi barrier design optimization, AC losses, T _CS performance, etc.). For Cu strand the main issue is RRR. For the TF conductor, the topics presented cover hydraulic correlation validation measurements (pressure drop tests), geometrical advanced controls, and electromagnetic behavior (hotspot checks and mainly the experimental results of the TF conductor full-size samples tested in SULTAN facility). The article presents comparisons of numerical simulation and experimental results used to confirm the design and the manufacturing processes.

10.1109/tasc.2012.2232951

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Citations (9)