[en] Superconducting coils in a Tokamak are subject to AC losses when the field transverse to the coil current varies. A simple model to evaluate the AC losses has been derived and benchmarked against a complete model used in the ITER design procedure. The influence of the feedback control strategy on the AC losses is examined using this model. An improved controller is proposed, based on this study. (author)

[en] Within the research program on the International Thermonuclear Experimental Reactor (ITER) Poloidal Field (PF) coils, a full size conductor sample was tested in the SULTAN facility (CRPP Villigen, Switzerland). This sample is composed of two straight ITER-like cable-in-conduit conductors, using the same NbTi strand. The two conductors are identical except for one leg makes use of a cable containing steel wraps around the main sub-cables as in the ITER design, while the other has no wraps inside. The paper presents conductor DC tests results compared to predictions given by various models developed within ITER-associated laboratories. These models aim to predict the DC behaviour of the cable from the experimental performances of the single strand. They have to explain the observed voltage-current (V-I) or voltage-temperature (V-T) characteristics, including the thermal runaways. The lower experimental performances compared to all expectations have shown the necessity to revise the models and to introduce a possible uneven current distribution among the strands of the cables. (authors)

[en] The design of the ITER magnet system is being finalized. The reference design of the winding pack of the TF coil is based on the use of circular conductors supported by radial plates. This design has been chosen for its high insulation reliability during operation. The overall TF coil structure includes pre-compression rings made of unidirectional fiber glass, which reduce the stress level in the outer intercoil structures and the coil case. The design of the central solenoid, including pre-load structure, has been developed. Two conductor jacket options are still under investigation for the CS and the final choice will be based on the results of on-going R and D

[en] Two kinds of anomaly, observed in the Volt-Ampere Characteristic (VAC) of large NbTi cable-in-conduit conductors (CICC) are discussed. In one case, the wavy behavior of the VAC close to the current sharing range is explained with oscillations of the cooling temperature of the order of 10-30 mK. A simulation of periodic temperature variations is done to reproduce the experimental behavior. In another case, voltage spikes in the VAC are correlated with saw-tooth signals from Hall sensors monitoring the conductor self-field, suggesting the occurrence of local quenches and recovery, with local current re-distribution. The analysis of the two kinds of anomaly, with their own signature on the VAC, provides valuable diagnostic tools for the interpretation of large size CICC test results

[en] A short sample of the NbTi cable-in-conduit conductor (CICC) manufactured for the ITER PF insert coil has been tested in the SULTAN facility at CRPP. The short sample consists of two paired conductor sections, identical except for the sub-cable and outer wraps, which have been removed from one of the sections before jacketing. The test program for conductor and joint includes DC performance, cyclic load and AC loss, with a large number of voltage taps and Hall sensors for current distribution. At high operating current, the DC behavior is well below expectations, with temperature margin lower than specified in the ITER design criteria. The conductor without wraps has higher tolerance to current unbalance. The joint resistance is by far higher than targeted. (authors)

[en] Theoretical predictions and tokamak experiments have shown that small deviations from the magnetic field axisymmetry, usually called error fields, may degrade the plasma performance in a large fusion device such as the ITER machine. The main causes of error fields are deviations from the ideal magnet shape due to tolerances. Estimates have been carried out to identify and quantify possible deviations in the coil system axisymmetry which may appear during coil manufacturing and machine assembly, particularly due to misalignment of the Toroidal Field (TF) and Poloidal Field (PF) coil systems. The expected error fields are calculated separately for each source of error. The results are used to form a basic (influence) matrix for the probabilistic combination of the error field from these sources. The results show that, with the advanced techniques selected to control the ITER magnet tolerances, the total level of error field is within the range which can be compensated by the correction coils included in the ITER design. (author)

[en] The ITER structures operating at cryogenic temperature suffer from heat deposition due to eddy current circulation. The events leading to eddy current generation have been categorized and analyzed for a typical ITER configuration and operational scenario. Prediction of the distribution of the energy deposition in the various cryogenic components is shown and commented. The effects of poloidal plasma current changes are also considered. (author)

[en] The Poloidal Field Conductor Insert (PFCI) of the International Thermonuclear Experimental Reactor (ITER) has been designed in Europe and is being manufactured at Tesla Engineering, UK, in the frame of a Task Agreement with the ITER International Team. Completion of the PFCI is expected at the beginning of 2005. Then, the coil shall be shipped to JAERI Naka, Japan, and inserted into the bore of the ITER Central Solenoid Model Coil, where it should be tested in 2005 to 2006. The PFCI consists of a NbTi dual-channel conductor, almost identical to the ITER PF1 and PF6 design, about 45 m long, with a 50 mm thick square stainless steel jacket, wound in a single-layer solenoid. It should carry up to 50 kA in a field of about 6 T, and it will be cooled by supercritical He at around 4.5 K and 0.6 MPa. An intermediate joint, representative of the ITER PF joints and located at relatively high field, will be an important new item in the test configuration with respect to the previous ITER Insert Coils. The PFCI will be fully instrumented with inductive and resistive heaters, as well as with voltage taps, Hall probes, pick-up coils, temperature sensors, pressure taps, strain and displacement sensors. The test program shall be aimed at DC and pulsed performance assessment of conductor and intermediate joint, AC loss measurement, stability and quench propagation, thermalhydraulic characterization. Here we give an overview of the preparatory work towards the test, including a review of the coil manufacturing and of the available instrumentation, a discussion of the most likely test program items, and a presentation of the supporting modeling and characterization work performed so far. (authors)

[en] The ITER magnets have been optimised and refined since the ITER Final Design Report (FDR) in 2001. Multiple design options have been eliminated and there is improved ability to drive a wide range of plasma configurations. Design iterations on the TF out of plane supports have eliminated stress concentrations in the inner keyways and have led to the choice of a so called friction-joint on the outside. The closure procedure for the TF case has been changed, with a new case segmentation, less risk of winding pack damage from shrinkage and better filling of the case-winding gaps. Selection of compact joints for the CS has enabled the peak field and cyclic stress levels in the conductor to be reduced while maintaining the flux capability. The uncertainty in the nuclear heat levels in the inner legs of the TF coils, and the need to operate with plasma nuclear powers from 360 to 700MW, lead to a thermal screen on the inside of the case with variable cooling capability. The electrical insulation specification has been refined after irradiation test results to give a better margin on the onset of degradation after operation to 3MWa/m². The RWM stabilisation provided by the side CC has been extended by accepting higher voltages and heating from AC losses. R and D results from the model coil tests have shown lower than expected design margins for the Nb3Sn conductors. This has been offset by adopting the latest advances in strand performance, and the margins of the new conductor will be confirmed by testing in 2005. Preparation for procurement is underway with considerations on technically acceptable ways of splitting the magnet supply. (author)

[en] The level of knowledge accumulated to date in the physics and technologies of controlled thermonuclear fusion (CTF) makes it possible to begin designing fusion—fission hybrid systems that would involve a fusion neutron source (FNS) and which would admit employment for the production of fissile materials and for the transmutation of spent nuclear fuel. Modern Russian strategies for CTF development plan the construction to 2023 of tokamak-based demonstration hybrid FNS for implementing steady-state plasma burning, testing hybrid blankets, and evolving nuclear technologies. Work on designing the DEMO-FNS facility is still in its infancy. The Efremov Institute began designing its magnet system and vacuum chamber, while the Kurchatov Institute developed plasma-physics design aspects and determined basic parameters of the facility. The major radius of the plasma in the DEMO-FNS facility is R = 2.75 m, while its minor radius is a = 1 m; the plasma elongation is k_9_5 = 2. The fusion power is P_F_U_S = 40 MW. The toroidal magnetic field on the plasma-filament axis is B_t_0 = 5 T. The plasma current is I_p = 5 MA. The application of superconductors in the magnet system permits drastically reducing the power consumed by its magnets but requires arranging a thick radiation shield between the plasma and magnet system. The central solenoid, toroidal-field coils, and poloidal-field coils are manufactured from, respectively, Nb_3Sn, NbTi and Nb_3Sn, and NbTi. The vacuum chamber is a double-wall vessel. The space between the walls manufactured from 316L austenitic steel is filled with an iron—water radiation shield (70% of stainless steel and 30% of water).