C. Jong

The ITER Magnet System

Procurement of the ITER magnets is due to start at the end of 2007/early 2008, with the launch of the longest lead time items, the conductor and the TF coil windings. The base design for procurement was established in 2001, and the build up of the Cadarache ITER team has been accompanied by a review of the most critical, or controversial, features of the 2001 design. At the same time, an urgent R&D program has been launched to complete the necessary verification of the design solutions that are proposed. In this paper an overview will be presented of the main design features and drivers, and some of the recent issues and R&D results will be summarized.

Overview of the ITER Correction Coils Design

The Correction Coils (CC) of the ITER Tokamak are developed to reduce the range of magnetic error fields created by imperfections in the location and geometry of the other coils used to confine, heat, and shape the plasma. The proposed system consists of three sets of 6 coils each, located at the top (TCC), side (SCC) and bottom (BCC) of the Tokamak device and using a NbTi cable-in-conduit superconducting conductor (CICC). Within each set, the coils are connected in pairs to produce a toroidal field to reduce the most troublesome, lower order, poloidal mode number fields (m = 1,2,3) in order to operate below the locked mode threshold. The conductor is designed to operate up to 6 T. The winding uses pancakes of one-in-hand conductor (quadpancakes for SCC, octopancakes for TCC and BCC), thus avoiding internal joints. The winding-pack is enclosed inside a 20 mm thick stainless steel casing. The coils are supported by rigid connections to the Toroidal Field (TF) coils. The structural design of the CC is mainly driven by the allowable fatigue stress levels in the conductor jacket, in the case material and in the glass-polyimide electrical insulation system. The boundary conditions on the CC are imposed by the TF coils deformation and the electromagnetic interactions with the PF coils system. The thermo-hydraulic and electrical performance of the CICC is also addressed.

Design and Specifications of the ITER TF Coils

The current design of the ITER Toroidal Field coils and structures, the main critical design and manufacturing issues, and the status of the procurement arrangements for these components, which will be released to the ITER parties in early 2008 to start the manufacturing contracts, are described. Some qualification and R&D work still required in preparation for the manufacture are also mentioned.

Progress in Design, Analysis, and Manufacturing Studies of the ITER Feeders

The feeder design has been improved by the feeder teams at the ITER Organization (IO) and the Institute of Plasma Physics, Chinese Academy of Science (ASIPP) by incorporating the results of mechanical and thermal analyses as well as the system integration and assembly tolerances in the present CAD model. The feeder design is being finalized progressively, and will be delivered to the Chinese Domestic Agent (CNDA) for further procurement arrangement (PA) activities. Pre-PA manufacturing studies and tests performed at ASIPP have been effective in clarifying feeder design feasibility and component manufacturability. This paper reports the recent advancements on feeder design, analysis and manufacturing studies.

Strain redistribution effects on current-sharing measurements on straight samples of large Nb3Sn cable-in-conduit conductors

Recent results on short samples of ITER-type Nb3Sn cable-in-conduit conductors have shown degradation, with the current-sharing temperature dropping slowly over several thousand current cycles. However, although such behaviour has been linked to the magnetic loads on the strands, which cause filament fracture and plastic yielding of the surrounding copper, the detailed examination of the results shows a number of inconsistencies. These suggest that the degradation may be exaggerated by artefacts of the short sample and may provide an over-conservative assessment of the in-coil performance. The behaviour of a short sample and its frictional interaction with the jacket have been analysed to consider the impact of local material modulus changes. Both an analytical approach and a finite-element simulation have been used. This paper shows that, if transverse magnetic loading causes a local reduction in the longitudinal elastic modulus of the cable in the high-field region, then the thermal compression along the cable becomes non-uniform, with a higher compression in the high field. The process develops with load cycles, due to the jacket/cable friction. If the strain change is interpreted using Jc, T, B and the strain scaling law, the predicted current-sharing temperature drop is similar to those observed during testing.

An Optimized Central Solenoid for ITER

The Central Solenoid (CS) of the ITER tokamak has to provide the flux variation needed to induce the plasma current and to shape the field lines in the divertor region. It is designed as a stack of 6 identical coils, independently power supplied. Repulsing forces arising between the coils during a scenario are withstood by a precompression structure installed around the coils. Studies were carried out to simplify the winding manufacture, to optimize the precompression structure and procedure, to optimize the stack assembly of the 6 coils and the assembly of the central solenoid inside the tokamak which allows withdrawal from the machine, while meeting the ITER design criteria and in particular the Magnet Structural Design Criteria (static and fatigue).

ITER Central Solenoid design

The Central Solenoid (CS) is a critical component in the ITER tokamak providing plasma current drive and shaping. The CS final design is being completed at the US ITER Project Office (USIPO) in Oak Ridge, TN under a Procurement Arrangement with the ITER Organization (IO). Key design decisions have been made and CAD models and drawings developed. Interfaces have been established. An extensive R&D program has been completed. Analyses have been conducted to verify the design meets requirements. Design documentation is being completed in anticipation of a Final Design Review in the fall of 2013. The paper describes the key features of the CS final design.

From Design to Development Phase of the ITER Correction Coils

The Correction Coils system (CC) within ITER, is intended to reduce the range of magnetic error fields created by assembly or geometrical imperfections of the other coils used to confine, heat, and shape the plasma. The proposed magnet system consists of three sets of 6 coils each, located at the top (TCC), side (SCC) and bottom (BCC) of the Tokamak device and uses a NbTi cable-in-conduit superconducting conductor (CICC) operating at 4.2 K. The ITER Organization (IO) and the Institute of Plasma Physics at the Chinese Academy of Sciences (ASIPP) are jointly preparing the definition of the technical specifications and the upcoming qualification program for the Correction Coils. The proposed design consists of a one in hand conductor winding without internal joint inserted in a structural casing which reacts the electromagnetic loads. The development of major items such as terminal joints, casing manufacture, and vacuum impregnation system, is an essential phase before the series production which will take place at the premises of the supplier. This paper discusses the key technologies on CC coils and future plans for short sample prototypes fabrication.

Moving Toward Manufacture of the ITER Central Solenoid

After several years of design optimization, the Central Solenoid (CS) of the ITER Magnet system is now moving towards manufacture. The design has evolved to take into account on one hand the results of the R&D carried out by the US ITER team in charge of the development of the design and on the other hand the feedback provided by the involvement of industry in preparation of the manufacture. To address specific issues, dedicated mock-ups have been manufactured and tested. Electromagnetic, structural and thermo-hydraulic analyses have been carried out to verify the compliance of the design with the ITER design criteria. A review of the Final Design is planned in 2013, preparing then to move into the manufacturing phase.

Addressing the Technical Challenges for the Construction of the ITER Central Solenoid

The Central Solenoid (CS) of the ITER Magnet system is split into six independently powered coils enclosed inside an external structure which provides vertical precompression thus preventing separation of the coils and acting as a support to net resulting loads. The analyses include an assessment of the mechanical behavior of the different components of the CS, under the normal and fault conditions, aiming at demonstrating the ability of the CS to achieve 30 000 cycles of plasma operation at nominal current (15 MA). A comprehensive material testing program is developed for the conductor jacket, the impregnated glass-epoxy insulation and the structure. The paper describes the architecture of the analysis and qualification programs and provides an overview of the results obtained so far.