Completion of the ITER toroidal field model coil

Maix, R.; Fillunger, H.; Hurd, F.; Salpietro, E.; Mitchell, N.; Libeyre, P.; Decool, P.; Ulbricht, A.; Zahn, G.; Corte, A. della; Ricci, M.; Bresson, Daniel; Bourquard, A.; Baudet, F; Schellong, B.; Theisen, E.; Valle, N.

doi:10.1016/s0920-3796(01)00420-3

Cited by 15 publications

(8 citation statements)

References 4 publications

Supporting

Mentioning

Contrasting

Order By: Relevance

“…Such features match very well with the requirements of conductors constituting nuclear fusion magnets and, as a matter of fact, most of the research and development in the field has occurred in this environment, where typical operating requirements for conductors are in the range: 20-70 kA current; 5-12 T peak magnetic fields. The most recent nuclear fusion devices based on the magnetic confinement concept, either already in operation (EAST [25][26][27], K-STAR [28,29]) or in construction phase (JT-60SA [30,31], ITER [32], SST-1 [33,34], W-7X [35,36]), all rely on the superconducting CICC technology. Without reviewing the whole history of CICCs for fusion, however, which would be largely a repetition of what previously reported in [19][20][21][22][23][24], we would like to highlight here the main design aspects that drive the performances of large-size CICCs for the specific requirements of fusion magnets, with special regard to the Nb 3 Sn conductor performance degradation with mechanical loading and fatigue.…”

Section: Main Design Aspects Of Cicc Technology: Nb 3 Sn Conductor Pe...mentioning

confidence: 99%

Cable-in-conduit conductors: lessons from the recent past for future developments with low and high temperature superconductors

Muzzi

Marzi

Zenobio

et al. 2015

Supercond. Sci. Technol.

View full text Add to dashboard Cite

We review progress in the design of high field superconducting cable-in-conduit conductors (CICCs) for fusion applications, with special attention to the results of recent key experiments, leading to the state-of-the-art CICC technology: the ITER Toroidal Field and Central Solenoid programs, the EFDA Dipole conductor development program, the NHFML Hybrid Magnet project, the EU-TF Alt conductor demonstration, and the CRPP React & Wind flat cable test. For these projects, the main CICC design driver was the mitigation of Nb 3 Sn conductor performance degradation with electro-magnetic loading cycles. This was achieved by proper choice of cable layout and of conductor geometry, depending on the specific operating conditions and project requirements. In all cases, the necessity to limit cable movements inside the conductor jacket was identified to be of crucial importance. The main aspects of CICC manufacture are also discussed here, at least for what is the experience gained by the authors in both CICC jacketing and cabling processes. Finally, the state of the art of high-temperature superconducting (HTS) cables is discussed: at present, this technology is still in its infancy, but it is highly likely that major technological improvements could eventually lead to a widespread use of HTS CICCs.

show abstract

Section: Main Design Aspects Of Cicc Technology: Nb 3 Sn Conductor Pe...mentioning

confidence: 99%

Cable-in-conduit conductors: lessons from the recent past for future developments with low and high temperature superconductors

Muzzi

Marzi

Zenobio

et al. 2015

Supercond. Sci. Technol.

View full text Add to dashboard Cite

show abstract

“…In the past, two insulation systems, commercially used for large area impregnation and applied as the turn and ground insulation for the toroidal field model coil (TFMC) [6] of ITER, were investigated in much detail. The results [7][8][9] demonstrated that their mechanical strength was seriously affected by radiation-induced damage.…”

Section: Introductionmentioning

confidence: 99%

Innovative insulation systems for superconducting fusion magnets

Humer¹,

Bittner-Rohrhofer²,

Fillunger³

et al. 2006

Supercond. Sci. Technol.

Self Cite

View full text Add to dashboard Cite

Glass fibre reinforced plastics (GFRPs) are usually employed as insulating materials for the superconducting coils of large fusion magnets, e.g. of the International Thermonuclear Experimental Reactor (ITER). Both the radiation spectrum and the stresses at the magnet location significantly influence the mechanical behaviour of the magnet insulation and, therefore, impose high demands on the material performance. During the last few decades, advanced epoxy based GFRPs with improved mechanical properties and radiation hardness were introduced into fusion technology. More recently, cyanate ester (CE) matrix systems have become of special interest. In this paper, various magnet insulation systems containing boron-free R-glass fibre reinforcements in commercial and new epoxies as well as in pure CE and CE/epoxy blended matrix systems are presented. All systems were irradiated in a fission reactor at ambient temperature (∼340 K) to a fast neutron fluence of 1 × 10 22 m −2 (E > 0.1 MeV). The mechanical properties were assessed at 77 K in tension as well as in interlaminar shear prior to and after irradiation under static and dynamic conditions.

show abstract

“…The ICS is also mechanically connected to the reinforced EURATOM LCT coil and is designed for an attractive load of 82.6 MN and an in-plane load of 14.6 MN. Details of the TFMC coil are given in [1]. Inside the vacuum vessel, the test assembly of the TFMC, the LCT and the ICS rest on 3 movable legs for carrying the weight of 108 t and compensating the thermal contraction during cool down.…”

Section: Introductionmentioning

confidence: 99%

Cryogenic Test Results of the ITER TF Model Coil Test in TOSKA

Zahn¹,

Bagnasco²,

Darweschsad³

et al. 2004

AIP Conference Proceedings

View full text Add to dashboard Cite

The ITER Toroidal Field Model Coil (TFMC) was designed and manufactured by the European Home Team in collaboration with European industry. The test in the TOSKA facility of the Forschungszentrum Karlsruhe was successfully performed in 2001 and 2002 and has confirmed that the used design and construction principles are applicable for the ITER TF coils.The TFMC was tested up to the rated current of 80 kA as a single coil and in the background field of the EURATOM LCT coil in order to achieve ITER TF coil relevant stress levels. For the operation of the TFMC and LCT coils, special developed forcedflow-cooled current leads were used. Both coils with a total weight of 108 t were forcedflow-cooled with supercritical He at 4.5 K in a secondary cooling loop connected to the 2 kW refrigerator. However, for currents above 11.4 kA in the LCT coil, its winding had to be cooled at 3.0 K with a separate refrigerator and cooling system.Details of the process engineering of both cooling systems will be described. The operation experiences during cool down, standby and current operation and recooling after fast discharges or Tcs measurements will be outlined hereafter.

show abstract

Completion of the ITER toroidal field model coil

Cited by 15 publications

References 4 publications

Cable-in-conduit conductors: lessons from the recent past for future developments with low and high temperature superconductors

Cable-in-conduit conductors: lessons from the recent past for future developments with low and high temperature superconductors

Innovative insulation systems for superconducting fusion magnets

Cryogenic Test Results of the ITER TF Model Coil Test in TOSKA

Contact Info

Product

Resources

About