In the ITER superconducting magnets, the CIC conductors cooling is insured by supercritical helium forced flow in the central and annular parallel channels. In the Central Solenoid, with a thick square jacket conductor, the helium inlet is in the highest field region at the inner bore pancake joggle, and it has to support the high hoop force stress level with a very low stress concentration factor. In the TF magnets, a thin jacket circular conductor is wound in double pancakes, inserted into radial plates, stacked and embedded into a steel case. The helium inlets are located at the inner bore, in the limited space between the radial plates and the coil casing. The PF coils helium inlets, not studied here, are similar to the CS ones, but with a lower stress level. A complete qualification work on the CS and TF helium inlets is presented. A design optimization was performed, by FEM analysis, resulting in acceptable stress level for both helium inlets. The welding procedure was qualified and specific fatigue life mock-ups were designed, analyzed and manufactured using representative jacket materials. Fatigue life qualification at 4 K was performed in the FZK test facility applying the relevant loading and number of cycles. Hydraulic mock-ups were manufactured and qualified in the CEA test facility using GN2 at relevant Reynolds number. Pressure drop as well as flow repartition inside the conductors' petals were measured. The mechanical and hydraulic results are presented
In parallel to the direct contribution to the procurement phase of ITER and Broader Approach, CEA has initiated research & development programmes, accompanied by experiments together with a significant modelling effort, aimed at ensuring robust operation, plasma performance, as well as mitigating the risks of the procurement phase. This overview reports the latest progress in both fusion science and technology including many areas, namely the mitigation of superconducting magnet quenches, disruption-generated runaway electrons, edge-localized modes (ELMs), the development of imaging surveillance, and heating and current drive systems for steady-state operation. The WEST (W Environment for Steady-state Tokamaks) project, turning Tore Supra into an actively cooled W-divertor platform open to the ITER partners and industries, is presented.
The Poloidal Field (PF) coils of the International Thermonuclear Experimental Reactor (ITER) are designed with NbTi cable-in-conduit conductors (CICCs) wound in double pancakes which are connected in series by joints. These joints have to operate in the poloidal magnetic field which generate large varying forces. The conceptual design of the joints is based on the overlap concept with twin-boxes and includes a tie-bar to carry the tensile load and a clamping support to resist the radial and vertical loads. The mechanical analysis is performed with several finite-element models of the joint area. It is shown that the design allows to keep both the tensile stress in the conductor and the shear stress in the insulation within acceptable limits.
The TF magnetic system of the international thermonuclear experimental reactor (ITER) will be composed of 18 superconducting coils assembled in a torus. The design of these large coils relies on the use of a circular cable-in-conduit conductor (CICC) composed of a cable of about one thousand twisted Nb 3 Sn and copper strands embedded in a thin steel jacket. The different thermal contractions of the materials composing this conductor during the cool down from the Nb 3 Sn heat treatment reaction temperature (about 920 K) to the operating temperature (about 5 K) result in a compressive strain on the superconducting Nb 3 Sn filaments which leads to a degradation of the conductor critical current. The qualification of the ITER conductor prototypes will be performed by testing short straight full size samples in the unique European test facility SULTAN at PSI (Switzerland). A comparison of the performance of the toroidal field model coil (TFMC) conductor in a short straight sample and in the coil has shown indications that the sample performance is better, apparently because the thermal strain in Nb 3 Sn is lower. It looks possible that this could have been caused by a slippage between the cable and the jacket in the joints at the sample ends during the heat treatment. In this case, these kind of short straight samples could not be relevant for conductor qualification. An investigation on the thermal strain by destructive examination of different TFMC conductor samples, including one leg of the TFMC short sample tested in 1999 in the SULTAN facility, was performed at CEA Cadarache. The paper reports on the different strain measurements, proposes an analytical approach of the results in order to conclude on the relevance of straight samples.
The strain sensitivity of Nb/sub 3/Sn cable is well known. However the practical process to compensate for this effect when 316 LN is used for the jacket has never been considered. In this paper different proposals are analysed in order to prevent the 316 LN jacket contracting more than the Nb/sub 3/Sn cable. A first experiment performed in the FBI test facility of KfK has shown that a prestrain of 0.3% carried out at 275 K on a short straight sample of cable in conduit conductor (3/spl times/3/spl times/4 Nb/sub 3/Sn strands of 0.73 mm in a 316 L conduit) produced an improvement of the critical current. The improvement in this condition is about 80%. Different designs of tooling usable for the CS and TF coils of ITER are described.
Superconductors for the ITER Poloidal Field Coils are large cable-in-conduit conductors (CICC) made of NbTi strands encased in a round-in-square stainless steel jacket. Three prototype conductor sections for poloidal field coils PF1/6, PF2/3/4 and PF5 have been fabricated in collaboration of the domestic RF, CN and EU agencies and tested in SULTAN Test Facility at the nominal operating conditions. The test aimed to characterize the DC and AC behavior of the conductors. The DC test was focused on the current sharing temperature (T cs ) at the nominal operating current and nominal operating background field. The take-off electric field at the nominal Helium mass flow rate was investigated versus the cable current density over a broad range of field and temperature. The AC loss measurement was performed before any electromagnetic loading and after a number of load cycles in order to define the impact of cyclic loads on the coupling currents constant of the cable. From the test results in SULTAN test facility, the margins in normal operation and the limits of the operation range of the ITER PF conductors are assessed.
The Central Solenoid (CS) is a key element of the ITER Magnet system, including six identical coils, called modules, assembled together to form a 4 m outer diameter, 13 m high solenoid. It is a superconducting magnet, using a 45 kA Nb3Sn conductor internally cooled by circulation of supercritical helium at 4.5 K with a peak field up to 13 T. It is enclosed inside a structure providing vertical pre-compression and mechanical support. Procurement of the components and the special assembly tooling of the ITER CS is the responsibility of US ITER, the ITER Domestic Agency of the USA, while the ITER Organization (IO) will carry out the assembly of these components. US ITER has awarded several contracts since 2011 to supply seven modules, including a spare, structure components, and the special tooling required for the CS pre-assembly. All deliveries are scheduled with the objective to start the CS assembly at IO site early 2021. IO is now actively preparing this new phase. This paper describes the general CS pre-assembly activities from modules stacking to the pre-compression at 210 MN. The special assembly processes and related tooling are detailed. A focus is given on the module lifting operation, the extensions assembly and the pre-compression. For these last two processes, IO has investigated alternative assembly options that will be presented.
The ITER Conductor Procurement Arrangement requires that all the conductors have to be qualified before the production phase. The objects of the work here presented are the Poloidal Field Coils conductors, specifically the PF1/6 and the PF2 conductors. The realization of the samples comprised the jacketing, the compaction and the straightening of the CICC, together with the design and manufacturing of the bottom hairpin box and the upper termination. The instrumentation and the related jacket machining completed the preparation of the samples. The samples have been assembled according to the specification defined by ITER and have been shipped to SULTAN facility. This paper describes all the activities performed during the preparation, the features of the samples and all the related issues.
Major steps have been completed in the manufacture of the ITER [Toroidal Field Model Coil] TFMC. We discuss experience gained in the construction of joints. The "Wind, React and Transfer" process and the radial plate and double pancake manufacture is presented.
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