Christelle Boyer

Development of the ITER PF Coils

The ITER Poloidal Field (PF) magnet system consists of six coils. Niobium-Titanium (NbTi) is used as superconducting material and cable-in-conduit conductor(CICC) type are used as a conductor. All coils are fabricated by stacking 6 to 9 double-pancakes wound by two-in-hand winding scheme. The six PF coils (PF1 to PF6) are attached to the Toroidal Field (TF) coil cases through flexible plates or sliding supports to allow small radial and vertical displacements. The outer diameters of the coils vary between 8 m and 24 m. Since the PF coil system provides magnetic field for plasma shaping and position control together with the Central Solenoid (CS) coil, it needs to operate in fast pulse mode, leading to induced voltages of up to 14 kV on the coil terminals during operation.

Validation of Helium Inlet Design for ITER Toroidal Field Coil

The ITER organization has performed design and its validation tests on a helium inlet structure for the ITER Toroidal Field (TF) coil under collaboration with CERN, KIT, and CEA-Cadarache. Detailed structural analysis was performed in order to optimize the weld shape. A fatigue resistant design on the fillet weld between the shell covers and the jacket is an important point on the helium inlet structure. A weld filler material was selected based on tensile test at liquid helium temperature after Nb3Sn reaction heat treatment. To validate the design of the weld joint, fatigue tests at 7 K were performed using heat-treated butt weld samples. A pressure drop measurement of a helium inlet mock-up was performed by using nitrogen gas at room temperature in order to confirm uniform flow distribution and pressure drop characteristic. These tests have validated the helium inlet design. Based on the validation, Japanese and European Union domestic agencies, which have responsibilities of the TF coil procurement, are preparing the helium inlet mock-up for a qualification test.

Design of the ITER PF Coil Joints

The ITER poloidal field magnet system consists of 6 pulsed coils with a diameter ranging from 8 m to 24 m and a weight of up to 400 t. The coils are made of independent modules connected to each other by joints. This paper summarizes the design constraints, including manufacturability and repair procedure, and the layout of the joints based on the “twin-box” concept. It particularly focuses on the structure of the joint and its support since the joints will operate at the maximum current of 55 kA in a 2 T magnetic field, and there- fore should be able to withstand high electromagnetic forces. In this article, we present the main results of the finite-element structural analyses of the joint. Finally, we summarize the main results of the electrical model developed to determine the losses (AC and transport current) and current sharing in the joint.

Computational Fluid Dynamics (CFD) Analysis of the Helium Inlet Mock-Up for the ITER TF Superconducting Magnets

The final design for the inlets of the supercritical He (SHe) coolant of the toroidal field (TF) superconducting magnets of the International Thermonuclear Experimental Reactor (ITER) has to be evaluated on the basis of different aspects, including the interest to minimize the associated localized pressure drop. Based on previous experience made on the analysis of the SHe inlets for the ITER superconducting central solenoid, we have developed and applied a computational fluid dynamics (CFD) model to compute the pressure drop versus mass flow rate characteristic in the TF inlet mock-up which was recently tested at CEA Cadarache, France. The cable model is first calibrated and verified against experimental data from short straight samples. The calibrated model is then applied to the inlet geometry and the results of the calculation are compared with the available measurements, showing very good agreement for sufficiently anisotropic permeability of the cable region. The thus validated model is finally used to investigate the distribution of the coolant flow among the different petals and the central channel, downstream of the inlet, as well as to quantitatively assess the role of the petal wrapping on the localized pressure drop at the inlet.

He-Inlet of the Toroidal Field Coil: Qualification and Manufacturing Status

In this paper, we will report on the manufacturing of 6 helium inlet mock-ups for the EU ITER TF coils, and on the results of the mock-up tests and other qualification activities carried out in the European industry on this subject.

Overview of the ITER Toroidal Field Magnet System Integration

The first series components of large D-shaped toroidal field coils (TFC) on the ITER Tokamak project are being fabricated and assembled at European Fusion for Energy (F4E) and Japanese Domestic Agency (JADA) premises since 2013. The TF magnet system consists of 18 individual coils connected in series based on a Nb3Sn cable-in-conduit conductors supplied by a 68-kA rated current with an overall 41-GJ stored energy and a peak magnetic field of 11.8 T. One of the key challenges of the construction of the 18 TFCs and their assembly resides in the control of the integration of the large individually manufactured coil components and in the ultimate management of tolerances on the final assembly into the Tokamak pit. This paper presents the integration aspects related to main TFCs subcomponents under fabrication starting from the TF conductor production, the winding of individual double pancakes, and their heat treatment and impregnation. This includes the fabrication of key prototypes for qualification purpose such as helium supply inlets, the electrical joints, and the design of the winding pack insertion into the structural TFC case during the final welding enclosure. Each preassembled 40° sector of a TFCs pair is then integrated into the torus according to tight tolerance requirements to provide both the so-called TF magnetic center line data and to guarantee the final operating wedged design into the inner leg region. The assembly of the coils terminal is then completed by connecting services through the power feeder busbars, the quench detection high voltage cables and the cryogenics interfaces pipe system.

Mechanical Fatigue Testing of a TF-He-Inlet Prototype at Cryogenic Temperature

During the superconducting magnet program for ITER, various cryogenic components need to be tested verifying their design. In the past a Helium inlet of the TF coil was tested under fatigue in 2005 following ASME criteria for fatigue qualification. However, this very first prototype achieved only 80% of the projected fatigue cycles. Looking at the breaking region it was concluded that an annealing of the HAZ took place during the welding process leading to a weakening of the metal possibly due to a poorly controlled weld.