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Abstract: The central solenoid (CS) of the ITER tokamak contributes to the inductive flux to drive the plasma, to the shaping of the field lines in the divertor region and to vertical stability control. It is made of 6 independent coils, using a Nb3Sn cable-in-conduit superconducting conductor, held together by a vertical precompression structure. This design enables ITER to access a wide operating window of plasma parameters, up to 17MA and covering inductive and non-inductive operation. Each coil is based on a stack of multiple pancake winding units to minimise joints. A glass–polyimide electrical insulation, impregnated with epoxy resin, is giving a high voltage operating capability, tested up to 29kV. The CS performance is fatigue driven mainly by the stress levels in the conductor jacket and in the precompression structure needed to keep the modules in contact during the repulsive forces which can arise in operation. A rigid connection to the TF coils provided at one end and a centering support at the other end allow to resist net vertical forces as well as unbalanced radial forces while avoiding torsion transmission from the TF Coils to the CS assembly. [Copyright &y& Elsevier]

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. [Copyright &y& Elsevier]

Abstract: The ITER magnets are a complex system involving interfaces between many advanced technologies (superconductors, forging/welding/machining of massive structures, cryogenics, composites and moulding, high voltage electrical), yet at the same time form part of the ITER ‘basic machine’ which is required to operate at the design parameters, broadly failure free, for the design life of the tokamak. This imposes special quality control problems for the ITER project integration by the ITER International Team (IT) through the design, fabrication and operation. The magnets are not a test bed for new technology but in spite of this must use it, successfully. There is little previous experience of such a system but full functionality is required from the start, with limited opportunity for adjustment. And, finally, costs and schedule must be contained. The procurement strategy for the machine, with magnet components being supplied ‘in kind’, requires particular attention to the specifications, scheduling and quality control (QC). Special issues here are the testing requirements on magnet components, especially before final installation but also at critical intermediate stages. Unnecessary or ineffective quality control procedures cause delay and high costs, and divert attention from critical items. The main points of the magnet QC programme are summarised, including the use of codes and standards, qualification, manufacturing quality assurance, commissioning and in-service inspection. [Copyright &y& Elsevier]

Testing of the five International Thermonuclear Experimental Reactor (ITER) model coils is nearly complete and it is possible to produce an overall summary of the conductor performance assessment. The interpretation is supported by a large database of tests on components (especially strand) and by analysis of the two main phenomena that influence the strand performance in the cable (current non-uniformity and mechanical loads). Individually, interpretation of individual coils is difficult but taken together the results enable conclusions to be drawn about the expected behaviour of conductors in the ITER coils. The conductor design has successfully resolved many of the problems of stability and pulsed current behaviour associated with large multi-strand cables, and the performance is better than expected in these areas. The current sharing behaviour is below expectations, probably due to an extra unexpected strain effect in the Nb3Sn strands arising from the local magnetic loads, and the paper concentrates on this issue. Design modifications are required to recover the performance and updated design criteria to implement these are proposed. [Copyright &y& Elsevier]

Abstract: A coherent set of Codes and Standards (C&S) for the ITER structures, system and components is needed to fulfil the safety and licensing requirements, and to provide the specified quality of the mechanical components manufactured by the different ITER Parties. In general the basic regulatory documents do not specify particular design and manufacturing codes, but for pressure equipment general Essential Safety Requirements are formulated and the means to fulfil them from a technical and legal point of view are given. This paper presents requirements for the ITER pressure and nuclear pressure equipment. It describes the selection of Codes and Standards for the mechanical components. Available industrial codes and standards (e.g. ASME, RCC-MR, and European standards) are selected for ITER components. For components not covered by conventional codes, specific structural design criteria and technical specifications have been developed. The remaining issues discussed are related to the completion of the specific C&S and the assessment of conformity with regulatory requirements before and during construction of the components. [ABSTRACT FROM AUTHOR]

The Central Solenoid (CS), a key component of the ITER Magnet system, using a 45 kA Nb 3 Sn conductor, includes six identical coils, called modules, to form a solenoid, enclosed inside a structure providing vertical pre-compression and mechanical support. Procurement of the components of the ITER CS is the responsibility of US ITER, the US Domestic Agency (USDA), while the assembly of these components will be carried out by the ITER Organization (IO). Procurement of all the coil modules was awarded in 2011 to General Atomics, while procurement of the structure is split among several manufacturers, using existing equipment, sometimes among the largest ones in the world. Assembly of the ITER CS will require a dedicated area in the ITER Assembly Hall, conventional tooling and special tooling. US ITER is in charge of the procurement of special tooling, while IO is responsible for the procurement of the conventional ones. A detailed assembly procedure is under development at US ITER, in close collaboration with IO and with the support of CEA. Procurement of the special Assembly Tooling is carried out by US ITER and the main part of the first item, the Assembly Platform, was delivered to IO in 2017. [ABSTRACT FROM AUTHOR]

Abstracrt ITER is incorporating two types of In-Vessel Coils (IVCs): ELM Coils to mitigate Edge Localized Modes and VS Coils to provide a reliable Vertical Stabilization of the plasma. Strong coupling with the plasma is required in order that the ELM and VS Coils can meet their performance requirements. Accordingly, the IVCs are mounted on the Vacuum Vessel (VV) inner wall, in close proximity to the plasma, just behind the Blanket Shield Modules (BSM). Due to high radiation environment, mineral insulated copper conductors enclosed in a stainless steel jacket have been selected. The reference design and prototype work provided a good basis for the development of radiation resistant conductor capable of operating within the harsh conditions in ITER vacuum chamber. However, this effort identified shortcomings in achieving satisfactory manufacturing solution, and most significantly, difficulties in brazing the brackets onto the ELM coil conductor. Since this process has not proven successful, alternative designs are under development and prototyping. Prototype manufacturing on the alternative designs has been completed at ICAS, Italy and ASIPP, China. The aim was to eliminate the need for internal coil joints, to prove the principle of longer conductor length manufacturing, and to perform bending and welding trials on two different conductor cross-sections: circular and square. The procurement of the IVCs and their conductors will be done via direct call for tender from the ITER Organization. This paper will give an overview of the alternative design and prototype manufacturing of the ITER In-Vessel coils. [ABSTRACT FROM AUTHOR]

This arc model was motivated by the need to model arcs during the failure of a large superconducting magnet following an unmitigated quench. During a quench, resistive heating raises the conductor and insulator temperature. Electrical and mechanical properties change and inline and bypass arcs can form. The arcs are sustained by the massive (gigaJoule) stored magnetic energy. As windings are bypassed by shorts, the inline arc current and hence, arc diameter, increases. Cable-in-conduit conductors limit the maximum arc column diameter and when limited, the arc properties change rapidly as the arc changes from a free arc to a confined arc. For arc current 10–100 kA we calculate the arc column electrical properties and temperature, by solving a set of equations describing the arc physics. The equations describe the arc column heating, gas ionisation, heat loss and electrical properties. By constraining the maximum arc column diameter in the solution, the transition between free and confined arc can be calculated. [ABSTRACT FROM AUTHOR]

The ITER Plasma Control System (PCS) approaching the second phase of development is now more deeply investigating alternative solutions for the various controls aimed at operations up to 15MA with low auxiliary heating in L-mode. The control functions in the PCS are strictly linked to the performance of the ITER actuators and diagnostics. The capabilities of those systems need to be carefully validated against the control needs. System and performance requirements shall be consistent with the control schemas and where limitations or restrictions are identified, it is necessary to provide alternative solutions. To guarantee this consistency, a set of interface documents is being prepared. Those interfaces for each of the plant systems that can impact the PCS activities detail the requirements specifically needed for control and report also the functional relationship between the two systems. The PCS has also to consider areas not actively part of plasma control that might affect or limit PCS operations (i.e. forces in the superconducting coils). This paper reports the main outcome from the interfaces definition. The actuator boundaries and plant systems constraints impacting the PCS design will be presented. For the sensors the challenge is the derivation of real-time measurement requirements in relation to the separate diagnostic requirements and their respective interface with the PCS. The complex organization of data integration with the PCS will be discussed. [ABSTRACT FROM AUTHOR]

Active mitigation of pulsed heat load is foreseen for the ITER superconducting magnets during cyclic plasma scenarios to ensure stable operation of the cryoplant. The paper addresses a promising feedback control to smooth heat pulses in the primary cryogenic circuits of the ITER magnets. A feedback control signal is proposed that can be derived from direct measurements of helium parameters in a low pressure return cryoline to the cryoplant. Configurations with two or more LHe baths were investigated at the Auxiliary Cold Box (ACB) level. Efficiency of the proposed control approach was evaluated in thermohydraulic simulations. A possibility to control several cryogenic loops using a common regulator is demonstrated. [ABSTRACT FROM AUTHOR]

A large amount of energy is stored in ITER superconducting magnet system. Faults which initiate a discharge are typically mitigated to quickly transfer away the stored magnetic energy for dissipation through a bank of resistors. In an extreme unlikely occurrence, an unmitigated fault event represents a potentially severe discharge of energy into the coils and the surrounding structure. A new simulation tool has been developed for the detailed study of these unmitigated fault events. The tool integrates: the propagation of multiple quench fronts initiated by an initial fault or by subsequent coil heating; the 3D convection and conduction of heat through the magnet structure; the 3D conduction of current and Ohmic heating both along the conductor and via alternate pathways generated by arcing or material melt. Arcs linking broken sections of conductor or separate turns are simulated with a new unconstrained arc model to balance electrical current paths and heat generation within the arc column in the multi-physics model. The influence under the high Lorenz forces present is taken into account. Simulation results for an unmitigated fault in a poloidal field coil are presented. [ABSTRACT FROM AUTHOR]

Abstract: The ITER correction coils (CC) include three sets of six coils each, distributed symmetrically around the tokamak and inserted between the toroidal field (TF) and the poloidal field (PF) coils. Each pair of coils located on opposite sides with respect to the plasma is series connected with polarity such to produce asymmetric fields. These superconducting coils use a cable-in-conduit conductor, insulated, wound into multiple pancakes and inserted inside an austenitic stainless steel case. The requirements and the main features of the design are presented and the selected options reviewed in terms of their criticality in achieving the specified tolerances. The requested qualification trials are identified and reports the results obtained so far. [Copyright &y& Elsevier]

Abstract: When energized, the ITER central solenoid (CS) coils experience large pulsed electromagnetic forces translating into a combination of tensile hoop load in the toroidal direction and transverse vertical compression to be resisted by the conductor jacket itself. A key parameter for the coil design is fatigue behaviour (60000 cycles) of the CS conductor jacket. The mechanical requirements on the CS conductor jacket are derived from the mechanical analyses performed on the CS stack assembly. A prototype production of the two considered candidate materials, 316LN and JK2LB, a specially developed high Mn-bearing austenitic stainless steel, has been achieved in industry. Manufacturing specifications and first results of production controls are presented. [Copyright &y& Elsevier]

Abstract: Error fields in Tokamaks are small departures of the exact axisymmetry of the ideal magnetic field configuration. Their reduction beyond a threshold value by the error field correction coils is essential since sufficiently large static error fields lead to discharge disruption. The error fields are originated not only by coils fabrication and installation alignment tolerances, joints and busbars but also due to the presence of ferromagnetic elements. The start of plasma current flattop with relatively low plasma density is considered as a critical state of the 15MA scenario for the onset of locked modes causing disruptions. A figure of merit, the ‘3-mode’ criterion, based on the lowest error field harmonics [(1,1); (2,1); (3,1)] was chosen to assess the error fields expected in ITER. Analysis performed last years in an independent way by CREATE (EU) and Efremov Inst. (RF) groups has allowed a deep understanding of the error fields induced by all the possible sources. Both groups were successfully benchmarked with the estimation of the error fields induced by a given deformed shape of a TF coil obtaining a perfect match of the results up to 2 orders of magnitude smaller than the defined threshold of B 3-mode/B to <5×10−5. Three different sets of independent variables based on a 3D rigid body movement of the coils have been provided for the 6 CS modules, 18 TF coils and 6 PF coils tolerances, which have allowed a clear understanding of the weight of each of the variables in the induced error fields. The results obtained in 2008 with a realistic set of magnets tolerances concluded that the system of correction coils provides effective suppression of the error fields with margin to correct possible impact of other sources. In particular, it has been shown that superconducting joints, feeders and busbars play a secondary effect; however the radial position of the TF coils and the tilt and radial shift of the CS stack would have a relevant influence. The ensuing recent sets of variables studied aimed at deepening the understanding of the operational limits. The present paper addresses the impact of the fabrication tolerances and installation misalignments of the magnet system; the impact of ferromagnetic inserts and busbars as well as the design of the correction coils will be thoroughly covered elsewhere (Amoskov et al., in press ). [Copyright &y& Elsevier]

Abstract: The ITER superconducting magnet system generates an average heat load of 23kW at 4K to the cryoplant, from nuclear and thermal radiation, conduction and electromagnetic heating, and requires current supplies 10–68kA to 48 individual coils. The helium flow to remove this heat, consisting of supercritical helium at pressures up to 1.0MPa and temperature between 4.3 and 4.7K, is distributed to the coils and structures through 30 separate feeder lines. The feeders also contain the electrical supplies to the coil, helium supply pipes and the instrumentation lines, and are integrated with the current lead transitions to room temperature. The components consist of the in-cryostat feeders, the cryostat feedthroughs and the coil terminal boxes (CTBs). This paper discusses the functional requirements on the feeder system and presents the latest design concept and parameters of the feeder components. [Copyright &y& Elsevier]

The Poloidal Field (PF) coils of ITER (International Thermonuclear Experimental Reactor) will supply the necessary magnetic field to initiate, shape, control and shutdown burning plasmas. The PF coils use NbTi cable-in-conduit superconductors, which operate at maximum currents of the order of 45–60 kA and experience large variations of current and magnetic fields. In order to test full-scale NbTi superconductors at operational conditions similar to ITER, the European Team has been asked to design and manufacture a PF conductor insert (PFCI). The cable has been provided by the Russian Federation. The Insert will be tested in 2004 in the Central Solenoid Model Coil (CSMC) facility at JAERI Naka, Japan. [Copyright &y& Elsevier]