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Journal Articles

Benchmarking of mechanical test facilities related to ITER CICC steel jackets

Vostner, A.*; Pong, I.*; Bessette, D.*; Devred, A.*; Sgobba, S.*; Jung, A.*; Weiss, K.-P.*; Jewell, M. C.*; Liu, S.*; Yu, W.*; et al.

IEEE Transactions on Applied Superconductivity, 23(3), p.9500705_1 - 9500705_5, 2013/06

 Times Cited Count:11 Percentile:51.85(Engineering, Electrical & Electronic)

The ITER Cable-In-Conduit Conductor (CICC) used in the superconducting magnet system consists of a cable made of 300 to 1440 strands housed in a stainless steel tube (a.k.a. jacket or conduit). There are circular, square, as well as circle-in-square jackets, made of either a very low carbon AISI 316LN grade stainless steel or a high Mn austenitic stainless steel developed for ITER called JK2LB. Selected mechanical properties of the base material and weld joint were tested at room temperature and/or cryogenic temperatures ($$<$$ 7 K). The Domestic Agencies (DAs) reference laboratories and the ITER-IO appointed reference laboratories, CERN and Karlsruhe Institute of Technology (KIT) performed mechanical tests. This paper will compare the test results (e.g. elongation to failure) from different laboratories.

Journal Articles

Preparation for the ITER central solenoid conductor manufacturing

Hamada, Kazuya; Nunoya, Yoshihiko; Isono, Takaaki; Takahashi, Yoshikazu; Kawano, Katsumi; Saito, Toru; Oshikiri, Masayuki; Uno, Yasuhiro; Koizumi, Norikiyo; Nakajima, Hideo; et al.

IEEE Transactions on Applied Superconductivity, 22(3), p.4203404_1 - 4203404_4, 2012/06

 Times Cited Count:17 Percentile:65.06(Engineering, Electrical & Electronic)

Japan Atomic Energy Agency (JAEA) has the responsibility for procurement of all of the ITER central solenoid (CS) conductor lengths. The CS conductor is composed of 576 Nb$$_{3}$$ Sn superconducting strands and 288 Cu strands assembled together into a multistage cable and protected by a circle-in-square sheath tube (jacket) with the outer dimension of 49 mm. In preparation for CS conductor production, the following R&D activities have been performed; (1) Mechanical tests at 4 K have been performed for jacket candidate materials such as 316LN and JK2LB, (2) Welding test for filler selection, (3) Measurement of coefficient of sliding friction using a 100-m long dummy cable, (4) Deformation characteristics of the conductor cross section after compaction and spooling. As a result of these R&D, the CS conductor jacket manufacturing technologies have been confirmed to start the procurement of the CS conductor.

Journal Articles

Status of ITER conductor development and production

Devred, A.*; Backbier, I.*; Bessette, D.*; Bevillard, G.*; Gardner, M.*; Jewell, M.*; Mitchell, N.*; Pong, I.*; Vostner, A.*

IEEE Transactions on Applied Superconductivity, 22(3), p.4804909_1 - 4804909_9, 2012/06

 Times Cited Count:125 Percentile:96.97(Engineering, Electrical & Electronic)

The ITER magnet system is made up of 4 sets of coils: 18 Toroidal Field (TF) coils, 6 Poloidal Field (PF) coils, 6 Central Solenoid (CS) coils and 9 pairs of Correction Coils (CC's). All of them are wound from Cable-In-Conduit Conductors (CICC's) made up of superconducting and copper strands assembled into a multistage, rope-type cable inserted into a conduit of butt-welded austenitic steel tubes. The TF and CS conductors call for about 500 tons of Nb$$_3$$Sn strands while the PF and CC conductors need around 250 tons of NbTi strands. The required amount of Nb$$_3$$Sn strands far exceeds pre-existing industrial capacity and calls for a significant worldwide production scale up. After recalling the technical requirements defined by the ITER Internal Organization (IO), we detail the in-kind procurement sharing of the various conductor types among the 6 ITER Domestic Agencies (DA's) involved: China, Europe, Japan, South Korea, Russia, and the United States, and we present a status of ongoing productions. The most advanced production is that for the TF coils, where all 6 DAs have qualified suppliers and have already registered more than 30% of the expected production data into the web-based ITER Conductor Database developed by the IO.

Journal Articles

Addressing the technical challenges for the construction of the ITER Central Solenoid

Libeyre, P.*; Bessette, D.*; Jewell, M.*; Jong, C.*; Lyraud, C.*; Rodriguez-Mateos, M.*; Hamada, Kazuya; Reiersen, W.*; Martovetsky, N.*; Rey, C.*; et al.

IEEE Transactions on Applied Superconductivity, 22(3), p.4201104_1 - 4201104_4, 2012/06

 Times Cited Count:7 Percentile:42.2(Engineering, Electrical & Electronic)

The Central Solenoid (CS) of the ITER magnet system will play a major role in tokamak operation, providing not only the major part of the inductive flux variation required to drive the plasma but also contributing to the shaping of the field lines and to vertical stability control. To meet these requirements, the design has been optimised by splitting the CS into six independently powered coils enclosed inside an external structure which provides vertical precompression thus preventing separation of the coils and, additionally, acting as a support to net resulting loads. To ensure that the CS design meets the ITER criteria, several analyses are performed along with a series of R&D trials to qualify the technologies to be used for the manufacture of the conductor, the coils and the structure.

Journal Articles

First qualification of ITER toroidal field coil conductor jacketing

Hamada, Kazuya; Takahashi, Yoshikazu; Isono, Takaaki; Nunoya, Yoshihiko; Matsui, Kunihiro; Kawano, Katsumi; Oshikiri, Masayuki; Tsutsumi, Fumiaki; Koizumi, Norikiyo; Nakajima, Hideo; et al.

Fusion Engineering and Design, 86(6-8), p.1506 - 1510, 2011/10

 Times Cited Count:11 Percentile:65.85(Nuclear Science & Technology)

Japan Atomic Energy Agency has a responsibility for procurement of the ITER toroidal field coil conductors as Japanese Domestic Agency (JADA) of the ITER project. The TF conductor is a circular shaped cable-in-conduit conductor, which is composed of cable and stainless steel conduit (jacket). The outer diameter and wall thickness of jacket are 43.7mm and 2mm, respectively. The cable consists of 900 Nb$$_{3}$$Sn superconducting strands and 522 Cu strands. The length of TF conductor is 780m in maximum. Preparation of conductor fabrication was completed in December 2009. And then, to demonstrate a conductor manufacturing procedure, JADA fabricated 780m-long Cu dummy conductor as a process qualification. Finally, the 780m-long Cu dummy conductor has been successfully completed, ahead of other domestic agencies that are in charge of TF conductor procurement. Since all of manufacturing processes have been qualified, JADA started to fabricate superconducting conductors for TF coils.

Journal Articles

Stability and quench analysis of toroidal field coils for ITER

Takahashi, Yoshikazu; Yoshida, Kiyoshi; Nabara, Yoshihiro; Edaya, Masahiro*; Bessette, D.*; Shatil, N.*; Mitchell, N.*

IEEE Transactions on Applied Superconductivity, 17(2), p.2426 - 2429, 2007/06

 Times Cited Count:14 Percentile:58.82(Engineering, Electrical & Electronic)

The ITER TF coils consists of 18 D-shape coils. The operating current, the maximum field and the stored magnetic energy are 68 kA, 11.8 T and 41 GJ, respectively. A Nb$$_{3}$$Sn cable-in-conduit conductor with a central channel is used, with a cooling length of 380 m. An accurate prediction of the coil performance requires, in addition to assessments of the superconductor behavior, a thermohydraulic analysis of the supercritical He. The overall thermohydraulic conditions were simulated by the full-scale quasi three dimensional code VINCENTA. Analysis of stability and quench was carried out using one dimensional Gandalf electric and thermohydraulic code. An interface was written between these codes. The stability margin against the mechanical disturbance and due to a plasma disruption was estimated. In the quench analysis, the temperature rise during the fast discharge was calculated. According to these results, it is confirmed that the TF coils will be operated with the designed performance.

Journal Articles

Performance of joints in the CS model coil and application to the full size ITER coils

Takahashi, Yoshikazu; Yoshida, Kiyoshi; Mitchell, N.*; Bessette, D.*; Nunoya, Yoshihiko; Matsui, Kunihiro; Koizumi, Norikiyo; Isono, Takaaki; Okuno, Kiyoshi

IEEE Transactions on Applied Superconductivity, 14(2), p.1410 - 1413, 2004/06

 Times Cited Count:10 Percentile:48.65(Engineering, Electrical & Electronic)

Cable-in-conduit conductors that consist of about 1,000 Nb$$_{3}$$Sn strands with an outer diameter of about 0.8mm, have been designed for the TF and CS coils of the ITER. The rated current of these coils is 40 -68kA. Two joint types (Butt and Lap) were developed during the CS Model Coil project. The performance of these joints was evaluated during the operating tests and the satisfied results were obtained. The joints of the TF coils are located outside of the winding in a region where the magnetic field is about 2.1T, a very low value as compared to the maximum field of 11.8T at the winding. The CS joints are located at the coil outer diameter and embedded within the winding pack due to the lack of the space. The maximum fields at the CS joint and winding are 3.5 and 13T, respectively. For the TF coils and the CS, the joints are cooled in series with the conductor at the outlet. The maximum temperature increase due to the joule heating in the joints is set at 0.15K to limit the heat load on the refrigerator. It is shown that both joint types are applicable to the ITER coils.

Journal Articles

Proposals for the final design of the ITER central solenoid

Yoshida, Kiyoshi; Takahashi, Yoshikazu; Mitchell, N.*; Bessette, D.*; Kubo, Hiroatsu*; Sugimoto, Makoto; Nunoya, Yoshihiko; Okuno, Kiyoshi

IEEE Transactions on Applied Superconductivity, 14(2), p.1405 - 1409, 2004/06

 Times Cited Count:16 Percentile:60.45(Engineering, Electrical & Electronic)

The ITER Central Solenoid (CS) is 12m high and 4m in diameter. The CS consists of a stack of 6electrically independent modules to allow control of plasma shape. The modules are compressed vertically by a pre-compression structure to maintain contact between modules. The CS conductor is CIC conductor with Nb$$_{3}$$Sn strands and a steel conduit. The CS model coil and insert coil test results have shown that the conductor design must be modified to achieve an operation margin. This required either to increase the cable diameter or to use strand with a higher current capability. A bronze-process (NbTi)$$_{3}$$Sn strand is proposed to achieve a higher critical magnetic field. A square conduit with a high Mn stainless steel is proposed as it can satisfy fatigue requirements. The inlets are in the high stress region and any stress intensification there must be minimized. The pre-compression structure is composed of 9tie plates to reduce the stress on the cooling pipes. These design proposals satisfy all ITER operational requirements.

Journal Articles

Key features of the ITER-FEAT magnet system

Okuno, Kiyoshi; Bessette, D.*; Ferrari, M.*; Huguet, M.*; Jong, C.*; Kitamura, Kazunori*; Krivchenkov, Y.*; Mitchell, N.*; Takigami, Hiroyuki*; Yoshida, Kiyoshi; et al.

Fusion Engineering and Design, 58-59, p.153 - 157, 2001/11

 Times Cited Count:2 Percentile:19.91(Nuclear Science & Technology)

no abstracts in English

Journal Articles

First test results for the ITER central solenoid model coil

Kato, Takashi; Tsuji, Hiroshi; Ando, Toshinari; Takahashi, Yoshikazu; Nakajima, Hideo; Sugimoto, Makoto; Isono, Takaaki; Koizumi, Norikiyo; Kawano, Katsumi; Oshikiri, Masayuki*; et al.

Fusion Engineering and Design, 56-57, p.59 - 70, 2001/10

 Times Cited Count:17 Percentile:75.42(Nuclear Science & Technology)

no abstracts in English

Journal Articles

Progress of the ITER central solenoid model coil programme

Tsuji, Hiroshi; Okuno, Kiyoshi*; Thome, R.*; Salpietro, E.*; Egorov, S. A.*; Martovetsky, N.*; Ricci, M.*; Zanino, R.*; Zahn, G.*; Martinez, A.*; et al.

Nuclear Fusion, 41(5), p.645 - 651, 2001/05

 Times Cited Count:55 Percentile:83.36(Physics, Fluids & Plasmas)

no abstracts in English

Oral presentation

Updating the design of the poloidal field coils for the ITER magnet system

Yoshida, Kiyoshi; Takahashi, Yoshikazu; Jong, C.*; Bessette, D.*; Mitchell, N.*

no journal, , 

The ITER magnet system consists of 18 TF coils, 6 PF coils, 6 CS modules, 18 CC and their feeders. Because of the difficulties in replacing the PF coils, the most unreliable component (the coil insulation) is designed with extra redundancy. There are two insulation layers with a thin metal screen in between. By monitoring the voltage of the intermediate screen, it is possible to detect an incipient short, defined as a short in only one of the two insulation layers. Alternately the faulty double pancake must be disconnected and by-passed because PF conductor uses NbTi strand. This implies that the remaining pancakes are operated at a higher current as the backup mode. Jumpers are pre-installed on the coil surface to allow this reconnection with a minimum of works in the cryostat and the conductor is designed with an extra margin. This paper presents the latest design for by-pass operation in PF coil, and thermal analysis of its conductor for the backup mode. The PF coil has an enough design margin for normal and backup mode.

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