Status of the ITER magnets

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Abstract

The first 2 years of the ITER IO has seen substantial progress towards the construction of the magnets, in three main areas. Firstly, the design has been developed under the conflicting constraints to minimise construction costs and to maximise plasma physics performance. Building construction momentum while updating the design to take account of new physics assessments of the coil requirements has been challenging. Secondly, with a stabilising design, it has been possible for the Domestic Agencies to launch the first industrial procurement contracts. And thirdly, critical R&D to confirm the performance of the Nb3Sn cable in conduit design is proceeding successfully.

The design consolidation has been accompanied by design reviews involving the international community. The reviews conducted by magnet experts have enabled a consensus to be built on choosing between some of the design options in the original ITER basic design in 2001. The major design decisions were to maintain the circular Nb3Sn conductor embedded in radial plates for the toroidal field (TF) coils and to maintain NbTi-based conductors for the PF coils. Cold testing, at low current, is also being introduced for quality control purposes for all coils.

Introduction

The ITER magnet system has been described recently in several papers [1]. The description is briefly repeated here in the interests of clarity. To avoid excessive duplication, the figures and explanation are directed at the integration issues of the magnets with the rest of the machine. We then describe the latest design progress, particularly in the area of tolerance definition, before focusing on final adjustments made to improve the machine plasma performance.

The ITER machine is contained in a pit, below ground level, as illustrated in Fig. 1. The ground level (of the assembly hall) is close to the bioshield on top of the cryostat dome and is the level from which components enter the tokamak building. The cutaway shows the galleries surrounding the main biological shielding around the cryostat, and the magnet CTBs (current leads, helium supplies and instrumentation) placed on two levels (upper and lower) outside the biological shield and connected to the coils through the feeders that lead into the cryostat. The magnet system itself consists of 18 toroidal field (TF) coils, a central solenoid (CS) with 6 individually powered modules, six poloidal field (PF) coils and 18 correction coils (CCs), Fig. 2. Supplies to the coils come from below or above and compete for space inside the cryostat with vacuum vessel supports, in-vessel access ducts and supply pipes for the vessel and contents. All coils are superconducting and the main parameters are summarised in Table 1.

The TF coils generate the field to confine charged particles in the plasma, the CS provide the inductive flux to ramp up plasma current and contribute to plasma shaping, the PF provide the position equilibrium of plasma current (i.e. the fields to confine the plasma pressure) and the plasma vertical stability. The CCs (see also Fig. 6) allow correction of error field harmonics (up to 3 harmonics in toroidal and poloidal directions) due to coil positioning and manufacturing errors as well as from busbars and feeders.

The TF coil case encloses the winding pack and is the main structural component of the magnet system, Fig. 3. The TF coil inboard legs are wedged all along their side walls in operation, with friction playing an important role in supporting the out-of-plane magnetic forces. In the curved regions above and below the inboard leg, the coils are structurally linked by means of three upper and three lower precompression rings formed from unidirectional bonded glass fibre that provide compression on four sets of upper and four lower poloidal shear keys arranged normal to the coil centreline. In the outboard region, the out-of-plane support is provided by four sets of outer intercoil structures (OIS) integrated with the TF coil cases and positioned around the perimeter within the constraints provided by the access ducts to the vacuum vessel. The OIS form four toroidal rings and act as shear panels in combination with the TF coil cases. There is low voltage electrical insulation toroidally between TF coils in the inboard leg wedged region, at the poloidal shear keys and between the OIS connecting elements.

The coil joints are located at the bottom of the coil just inboard of the gravity support. The winding consists of double pancakes (DPs), with the helium cooling inlets on the inner cross-over (opposite the joint region in the low field part of the coil) and all the joints and manifolding are contained in a local case extension below the coil.

The gravity supports are composed of pedestals (one under each TF coil), with flexible elements to allow radial displacements during cool-down and operation that would otherwise create extra stresses in the cases.

The main features of the TF windings are given in Table 2.

The CS assembly consists of a vertical stack of six independent winding pack modules, which is hung from the top of the TF coils through its pre-load structure, Fig. 4 and [2]. At the bottom, there is a sliding connection to provide a locating mechanism and support against dynamic horizontal forces. This pre-load structure, which consists of a set of tie-plates located at the inner and outer diameters of the coil stack, provides axial pressure on the stack. The modules can be energised independently and busbars and joints are placed outside the coils, with helium supply and return lines inside the bore. The CS stack is self-supporting against the coil radial forces and most of the vertical forces, with the support to the TF coils reacting only the weight and net vertical components resulting from up-down asymmetry of the poloidal field configuration. The CS pre-load structure consists of the lower key block, the upper key block, a set of 9 inner and 9 outer tie-plates, buffer plates, and connecting bolts. The flanges are split into 9 sectors linked by electrically insulated bolted joints to reduce AC losses during pulsing of the machine and to allow the structure to flex with the radial expansion of the top and bottom coils (avoiding shear and slip at the interface). The outside tie-plate region needs to have at least 30% open space in the toroidal direction for the joints, current leads and helium pipe arrangement (Fig. 5).

The individual coil modules are identical but rotated and inverted to form the stack, with interface plates bonded to each during stacking. Each winding is formed from six hexapancakes and one quadrapancake, with helium inlets at the inner (high field) cross-overs and all joints/helium outlets on the outside, Table 2.

The six PF coils (PF1 to PF6) are attached to the outside of the TF coil cases through flexible plates or sliding supports allowing radial displacements, Fig. 2. The PF coils positions and sizes have been optimised for the plasma requirements, within the constraints imposed by the access and pumping ducts to the in-vessel components.

The winding uses double pancakes, with joints only on the outside for space and eddy current/AC loss reasons, and helium inlets at the inner cross-overs.

Outside the TF coils are located three independent sets of CCs, each consisting of six coils arranged around the toroidal circumference above, at and below the equator, Fig. 6. Within each set, pairs of coils on opposite sides of the machine are connected in series. These coils are used to correct error fields (particularly toroidal asymmetry) from the busbars and positioning errors in the TF coils, CS and PF coils. They can also correct error fields from the neutral beam systems. They may in addition be used for feedback control of plasma resistive wall modes (RWM).

Both CS and TF coils operate at high field and use Nb3Sn-type superconductor. The PF coils and CCs use NbTi superconductor. All coils are cooled with supercritical helium with a coil inlet temperature of 4.5 K. The conductor is a cable-in-conduit conductor with a circular multistage cable consisting of about 1000 strands cabled around a small central cooling spiral tube. The cable is contained in a circular jacket for the TF coils or a jacket with an outer square section for the other coils. The operating currents are 40–45 kA for the CS, 45–55 kA for the PF coils and 68 kA for the TF coils. The CCs use a reduced size conductor, 16 kA, with about 300 strands and without the central cooling channel.

Table 3 summarises the coil voltages [3]. These voltages are the maxima in normal conditions (which of course must include the effect of ripple in the DC supply and the variation between on-load and off-load output from the converters) and include the extra capability required for upgrades to the vertical stabilisation control circuits [4]. In addition to these voltage levels, there is a requirement for the voltage to resist fault conditions, within the power supplies, control system or electrical circuits (up to the level of a single short). This brings required test voltages up to 29 kV for CS and PF coils and 20 kV for the TF. The coil electrical insulation system is composed of multiple layers of polyimide film-glass impregnated with epoxy resin. Epoxy-glass is used extensively to fill tolerance gaps. The CS, PF and CC coils are pancake wound with a conductor that has a square outer section. The well-known insulation weakness with square conductors which creates insulation stress concentrations at the corners under compression conditions is offset in the PF and CS coils with a thick epoxy-glass layer as a cushion and a multiple overlapped polyimide-glass sandwich providing the electrical strength. The TF coils use a conductor with a circular outer section that is contained in grooves in so-called “radial plates” which provide a uniform circular turn insulation with low stresses. There is one radial plate for each double pancake and the conductor is contained in grooves on each side, Fig. 4.

Each coil is connected through an in-cryostat feeder to the current leads and pneumatic coolant valves located just outside the main cryostat in the coil terminal boxes (CTBs). Each feeder has a robust steel outer containment that provides protection to the cryostat components in the event of a short within the current supply busbars. It consists of the feed and return current supply busbars (using NbTi superconductor), the return and supply helium lines and instrumentation lines. The busbars themselves have an outer grounded steel shell which contains a thick vacuum impregnated layer of epoxy-glass–polyimide insulation. Each pair of superconducting busbars is separated by a steel plate which extends along the entire length of the conduit, providing a temporary low resistance ground path as additional protection against a short circuit between busbars. The necessary flexibility for thermal and mechanical movement of the coils is obtained by the use of S bends in the busbars and piping.

ITER has a total of 60 current leads using high temperature superconductor (HTS), The TF coils are connected in pairs with an internal s/c busbar (substantially reducing the space requirement in and around the cryostat) and all other coils are connected individually.

A similar concept is used also for the cooling of the structures, with the feeders connected to coil valve boxes (CVBs).

Section snippets

Tolerances

One of the most critical tasks before procurement of the coils can start is the definition of the tolerances of the components. The final magnet tolerances are driven by three considerations, firstly the allowable error fields (the radial and vertical field harmonics in the toroidal direction and toroidal harmonics in the poloidal direction that can substantially degrade plasma confinement) [5], secondly the need for components to fit together on assembly [6] and thirdly the allowable increase

Modifications to improve plasma performance

Following the start of the ITER construction activity, multiple design reviews have been undertaken of the magnet design and main manufacturing issues. Overall, technical reviews of the magnet design have had a small impact, demonstrating broad consensus among the ITER partners. Some simplification of design details has been carried out for manufacturing purposes. The largest proposed modification concerns the coil quality control programme. Full coil acceptance testing was originally proposed

R&D

Magnet R&D is now transitioning to prototype testing and detailed design performance qualification, rather than more basic R&D. Five of the most significant activities underway in 2007–2008 have been picked out here for a more detailed description.

Procurement

As described elsewhere [1] the magnets are largely an in-kind procurement, with the bulk (about 70% by value) supplied by China, Europe and Japan. Sharing of the components was agreed during the period of ITER negotiations in 2003–2005 and can now only be changed with difficulty, requiring the agreement of all ITER partners. The basis for the procurement is a ‘build to print’ specification from ITER. This is not a full manufacturing design, but a design that is sufficiently detailed (and

Conclusions

Technical work on the magnets in the IO itself has made substantial progress. Working closely with the DAs, almost all coil components are now described by adequate drawings that include reasonable tolerances and integration of the auxiliary systems (particularly feeders and current leads). Interfaces definition, especially to plasma engineering, power supplies and cryoplant, is nearing completion. Analysis to support the design functionality, mainly on structures and thermohydraulics, is

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