Search

ITER In-Vessel Coils (IVCs), resistive magnets to be installed in close proximity to ITER plasma to compensate fast perturbations of the plasma itself, have undergone a comprehensive revision of their requirements and operational loads, leading to design modifications and R&D activities. An update of the plasma operating scenarios has been done and the maximum currents during transient plasma events have been assessed considering actual operating currents and the surrounding vacuum vessel support structure leading to more representative load cases. The main IVC component modifications are the conductor material and the winding pack support structure. Phase 1 of the conductor procurement involving two suppliers has been completed and the supplier selection for the main production is ongoing. Additional modifications include the joints, where Cu brazing has been replaced by welding, and the conductors are not brazed to the winding pack bracket but clamped. Bracket mockups for the coils have been completed identifying welding parameters and corresponding non-destructive examination (NDE), while different designs of electrical insulating breaks (IB) using Alumina have been tested against pressures up to 80 bars on top of the electrical characterization. A summary of the key challenges experienced towards design finalization is presented. [ABSTRACT FROM AUTHOR]

Abstract: Strands relevant for fusion with high critical current densities and moderate hysteresis losses were developed and already produced on industrial scale. Based on these achievements EFDA-CSU Garching has launched a Nb3Sn strand development and procurement action inside Europe in order to assess the current status of the Nb3Sn strand production capability. All six addressed companies have replied positively to the strand R&D programme which includes the three major Nb3Sn production techniques namely the bronze, internal-tin and powder-in-tube (PIT) route. According to the strand requirements for the ITER TF conductor a critical current density of 800A/mm2 (at 12T, 4.2K and 10μV/m) and overall strand hysteresis losses below 500kJ/m3 have been specified as the minimum guaranteed strand performance. The second major objective of this programme is to motivate the strand manufacturers to develop and design high performance Nb3Sn strands optimised for the ITER conductor. For this purpose, a target critical current density of 1100A/mm2 has been added to the specification. This paper describes the strategy behind the strand development programme, the actual status of the strand production as well as first preliminary results obtained from the strand suppliers. [Copyright &y& Elsevier]

The ITER in-vessel Vertical Stabilization (VS) coils, composed by the upper and lower VS coils, are designed to provide fast vertical stabilization of the plasma through a fast-current noise profile to compensate small plasma oscillations and to recover vertical displacement events. The VS coils are made by 4 continuous turns of Mineral Insulated Conductor (MIC) joined by Inconel ® 625 welded brackets, installed on rails on the inner wall of ITER Vacuum Vessel (VV). Thus, the coil turns winding/forming and the entire coils assembly have to be performed inside ITER VV, forcing the assembly procedure and tooling design to cope with the very limited space and strict constraints of ITER VV environment. The proposed procedure foresees that MIC conductor is un-spooled in the Neutral Beam (NB) port area and inserted into the ITER VV through a NB equatorial port, then a set of tooling is installed on a reinforced staging platform to wind, form and seal the conductor turns and finally to weld brackets under pre-compression. For this purpose, the tooling design including the reinforcement staging has been developed to meet the required tooling function, configuration optimization, assembly procedure and environmental constraints. The dedicated tools to be used for the assembly procedure of the VS coils are categorized into ex-vessel tools for unspooling and cleaning, in-vessel tools for winding and bump forming, bracket assembly tools for pre-compression and welding, termination and sealing tools, ancillary lifting tools and reinforcement staging platforms. This paper provides the overall assembly procedure and tooling development for the in-situ winding of VS coils. [ABSTRACT FROM AUTHOR]

• Feeders are the life lines of the ITER magnet system. • The PF4 CFT is the first ITER magnet component to be installed in the tokamak pit. • The Manufacture Inspection Plan (MIP) guarantees a good proficiency of the process of the Receiving, Inspection and Tests (RIT) phase. • Preliminary checks or trials are fundamentals avoiding potential future problems and saving time in the pit in the tokamak building. The ITER Magnets are in the procurement and assembly phases. During these phases, critical components need to be tested and assembly procedures need to be developed and qualified on mockups. To this end, ITER Organization (IO) and the Commissariat à l'Energie Atomique (CEA) have built up a support structure with space, expertise and equipment: MIFI – Magnet Infrastructure Facilities for ITER. In this framework, IO and CEA perform Receiving Inspection and Tests (RIT) on components delivered to IO by the Domestic Agencies (DA). RIT is a critical stage in the life process of components. These tests are meant to ensure the good quality of components and occur between the Factory Acceptance Test (performed before delivery to IO) and the Site Acceptance Tests that are performed after installation in the tokamak pit. The paper explains the RIT sequence managed on the Poloidal Field #4 Cryostat Feedthrough (PF4 CFT), the first ITER Magnets component to be delivered by a Domestic Agency to ITER Organization, and details leak tests, metrology tests and High Voltage tests performed. [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]

Abstract: Fusion for Energy (F4E) will manage the in-kind contribution from Europe (EU) to ITER of about one quarter of the machine magnet components. This contribution will consist in 10 Toroidal Field (TF) coils, 5 (out of 6) Poloidal Field (PF) coils, 20% of the Nb3Sn conductor used in the TF coils and 15% of the PF coil NbTi conductor. In addition, 12 pre-compression fibreglass rings for the TF system will be procured. This paper presents the complex organization and status of the procurement activities of the F4E magnet components. The technical and managerial challenges, in terms of fabrication, quality assurance and schedule, are also highlighted. [Copyright &y& Elsevier]

Abstract: In the framework of the European Fusion Programme, the Forschungszentrum Karlsruhe and the Centre de Recherches en Physique des Plasmas (CRPP) have successfully developed and tested a 70kA high-temperature superconductor (HTS) current lead demonstrator. In the present work, the optimisation of the cooling conditions with respect to the cooling power consumption and the required amount of HTS material is described for a binary current lead consisting of a conduction-cooled HTS part and a gas-cooled heat exchanger. The refrigerator input power, required to cool a binary HTS current lead, is determined by the heat load at the 4.5K level, the helium mass flow rate necessary to cool the heat exchanger part and the helium inlet temperature (T He,in). For a range of copper (T Cu) and helium inlet (T He,in) temperatures at the cold end of the heat exchanger, the optimum values of the length of the heat exchanger and the required helium mass flow rate have been calculated. Furthermore, the main results of the pulsed operation of the 70kA HTS current lead demonstrator are presented. The possibility of pulsed operation of HTS current leads for the CS and PF coils is discussed. [Copyright &y& Elsevier]

Abstract: The use of high temperature superconductor (HTS) materials in future fusion machines could increase the efficiency drastically, but strong boundary conditions exist. To outline the prospects, challenges and problems, first the benefit of using HTS materials is estimated considering the saving in cryogenic power. Next, it is demonstrated that industrial available HTS materials can be used for fusion today. For this purpose, we give a short summary of results that have been obtained from an ITER conform 70kA HTS current lead that was designed, built and tested by the Forschungszentrum Karlsruhe and the CRPP Villigen in the frame of the European Fusion Technology Programme and in cooperation with industry. This current lead consists of an HTS part that covered the temperature range from 4.5 to 70K and a conventional part, making the connection to room temperature. Because the HTS part had no ohmic losses and poor thermal conduction, the refrigerator power necessary for cooling the current lead was reduced drastically. The saving factor could be calculated to be 5.4 at zero current and 3.7 at 68kA. The current lead could even be operated at 80kA and with respect to safety criteria of ITER, a complete loss of He flow was simulated showing that the HTS current lead could hold a current of 68kA for 6min without active cooling. These results demonstrate that today existing HTS materials can be used in ITER for current leads or bus bar systems. For fusion machines beyond ITER, the development of an HTS fusion conductor would be the key to operate the complete magnet system at higher temperatures. The option of developing fusion conductors based on Bi-2223 and YBCO are briefly discussed. For a success of such conductors, the AC loss optimisation is crucial. [Copyright &y& Elsevier]

• The designed gap between IVC bolting system to overcome effects from thermal expansion is reasonable. • Thermal and mechanical cycling brings no obvious effects on the IVC insulation layer. • Long time action of EM load may make the fixed bolts loose. ITER In-vessel coils (IVCs) are the components to be installed in the vacuum vessel for plasma positions compensation controlling. Every coil will be wound continuously and clamped through the welded Inconel 625 brackets. The coils are planned to be bolted to the rails which are welded to the vacuum vessel wall behind the blanket shield modules. The operation condition is severe for the IVC coils, combined with high radiation, high temperature, as well as the high thermal and electromagnetic loads. To make sure the safe operation, tests have been done on the mock-up coil to check its fatigue properties under the combined thermal and mechanical loads before the finalization of the design. The integrity of the mock-up coil has been evaluated after the fatigue cycling tests. In this paper, the details are presented on the test procedures and the measurement results are summarized. [ABSTRACT FROM AUTHOR]

The coils inside the ITER vacuum vessel are part of the overall plasma control system that ensures and maintains stable plasma operations. Two vertical stability coils(VS)will provide fast vertical stabilization of the plasma, while an array of 27 edge localized mode (ELM) coils will create resonant magnetic perturbations to magnetically control the plasma exterior in order to suppress potentially harmful power deposition on plasma-facing components. For magnet coils operating inside of the vacuum vessel, conventional insulation schemes are not an option. After a decade of development and intensive collaboration with industry and research institutes, a copper mineral insulated conductor (MIC) in stainless steel jacket will be used as for its ability to withstand large transient electromagnetic forces, high radiation flux, and high temperature. Complex technologies have been developed by ASIPP for the series of production of IVC conductors, in terms of conductor initial dimension design, MgO block assembly and elaboration of Cu pipe insertion, compaction and winding processes. China has finished all qualification phases, especially for 100-meter unit length conductor manufacturing. The first 100-meter unit length of copper tube is successfully manufactured in China. All conductor samples required for qualification process have successfully passed the electrical tests with good performance. The challenges and status of the production of ITER IVC conductors in China are described in this paper. [ABSTRACT FROM AUTHOR]