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.Index Terms-Cable-in-conduit conductor, ITER, joint resistance, niobium-titanium, self-field induced quench.

The water-cooled lithium-lead breeding blanket is in the pre-conceptual design phase. It is a candidate option for European DEMO nuclear fusion reactor. This breeding blanket concept relies on the liquid lithium-lead as breeder-multiplier, pressurized water as coolant and EUROFER as structural material. Current design is based on DEMO 2017 specifications. Two separate water systems are in charge of cooling the first wall and the breeding zone: thermo-dynamic cycle is 295-328°C at 15.5 MPa. The breeder enters and exits from the breeding zone at 330°C. Cornerstones of the design are the single module segment approach and the water manifold between the breeding blanket box and the back supporting structure. This plate with a thickness of 100mm supports the breeding blanket and is attached to the vacuum vessel. It is in charge to withstand the loads due to normal operation and selected postulated initiating events. Rationale and progresses of the design are presented and substantiated by engineering evaluations and analyses. Water and lithium lead manifolds are designed and integrated with the two consistent primary heat transport systems, based on a reliable pressurized water reactor operating experience, and six lithium lead systems. Open issues, areas of research and development needs are finally pointed out.

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Since the installation of an ITER-like wall, the JET programme has focused on the consolidation of ITER design choices and the preparation for ITER operation, with a specific emphasis given to the bulk tungsten melt experiment, which has been crucial for the final decision on the material choice for the day-one tungsten divertor in ITER. Integrated scenarios have been progressed with the re-establishment of long-pulse, high-confinement H-modes by optimizing the magnetic configuration and the use of ICRH to avoid tungsten impurity accumulation. Stationary discharges with detached divertor conditions and small edge localized modes have been demonstrated by nitrogen seeding. The differences in confinement and pedestal behaviour before and after the ITER-like wall installation have been better characterized towards the development of high fusion yield scenarios in DT. Post-mortem analyses of the plasma-facing components have confirmed the previously reported low fuel retention obtained by gas balance and shown that the pattern of deposition within the divertor has changed significantly with respect to the JET carbon wall campaigns due to the absence of thermally activated chemical erosion of beryllium in contrast to carbon. Transport to remote areas is almost absent and two orders of magnitude less material is found in the divertor.

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The full set of T CS measurements performed during 2000-2002 on conductor 1A of the ITER Central Solenoid Model Coil (CSMC) of the International Thermonuclear Experimental Reactor (ITER) is analysed with the extensively validated M&M code. Under the assumptions of uniform strand properties and uniform current distribution among strands, the performance of this ''average'' strand in the cable is deduced from the best fit of the measured voltage-inlet temperature characteristics. Two fitting parameters are chosen: an ad-hoc additional contribution Ôe extra Õ to the longitudinal strain of the average strand in the cable, and the average-strand (or cable ''effective'') index ÔnÕ of the electric field-current density characteristic. It is shown that the average strand in the CSMC performed less well than expected from the strand database--an increasingly negative e extra being needed to reproduce the coil behaviour at increasing transport currents, and that the cable effective n is clearly below the measured n of the strand. It is argued that this average-strand performance reduction in the CSMC is most likely related to mechanical load effects.

The world's largest pulsed superconducting coil was successfully tested by charging up to 13 T and 46 kA with a stored energy of 640 MJ. The ITER central solenoid (CS) model coil and CS insert coil were developed and fabricated through an international collaboration, and their cooldown and charging tests were successfully carried out by international test and operation teams. In pulsed charging tests, where the original goal was 0.4 T/s up to 13 T, the CS model coil and the CS insert coil achieved ramp rates to 13 T of 0.6 T/s and 1.2 T/s, respectively. In addition, the CS insert coil was charged and discharged 10 003 times in the 13 T background field of the CS model coil and no degradation of the operational temperature margin directly coming from this cyclic operation was observed. These test results fulfilled all the goals of CS model coil development by confirming the validity of the engineering design and demonstrating that the ITER coils can now be constructed with confidence.