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Remarkable progress in the physical parameters of net-current free plasmas has been made in the Large Helical Device (LHD) since the last Fusion Energy Conference in Chengdu, 2006 (O.Motojima et al., Nucl. Fusion 47 (2007. The beta value reached 5 % and a high beta state beyond 4.5% from the diamagnetic measurement has been maintained for longer than 100 times the energy confinement time. The density and temperature regimes also have been extended. The central density has exceeded 1.0×10 21 m -3 due to the formation of an Internal Diffusion Barrier (IDB). The ion temperature has reached 6.8 keV at the density of 2×10 19 m -3 , which is associated with the suppression of ion heat conduction loss. Although these parameters have been obtained in separated discharges, each fusion-reactor relevant parameter has elucidated the potential of net-current free heliotron plasmas. Diversified studies in recent LHD experiments are reviewed in this paper.

The voltage-temperature characteristic curve (V-T curve) observed in the large-current Nb 3 Sn CIC conductor, which was used in the ITER CS Insert, showed more gradual take-off toward normal state than that of an individual strand composing the conductor. More gradual take-off corresponds to a reduction in so-called "n-value", and measured n-values of the strand and conductor of the CS Insert were 30 and 7, respectively. This reduction cannot be explained by a tensile strain of the conductor caused by a hoop deformation which is uniform along the conductor length. Investigation is therefore required to clarify the strain states of each strand, especially those caused by a transverse electromagnetic force acting on each strand. In a CIC conductor, since strands are twisted to form a cable, each strand is mechanically supported by nearby strands at an interval (typically 5 mm) set by the twist pitch. Between two supporting points, the strand is free to move under the transverse force and a cyclic deformation may occur along the strand length. This deformation will produce nonuniform bending strain along the strand. In order to verify the above consideration and to quantitatively evaluate the effect of this deformation, we prepared an apparatus to simulate this cyclic deformation by artificially applying a transverse load on the strand and its V-T characteristic was measured. When the strand received the transverse force of 10 3 10 4 N m (which is expected value for a strand of the CS Insert operated at 13 T, 46 kA), n-value reduced to less than 15 from the original value of 30, which agreed to the phenomena observed in the CS Insert. This indicates that the transverse force acting on each strand causes the reduction in n-value of the CIC conductor.

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.

The Central Solenoid Model Coil (CSMC) was designed and built by ITER collaboration between the European Union, Japan, Russian Federation and the United States in 1993-2001. Three heavily instrumented insert coils have been also built for testing in the background field of the CSMC to cover a wide operational space. The TF Insert was designed and built by the Russian Federation to simulate the conductor performance under the ITER TF coil conditions. The TF Insert Coil was tested in the CSMC Test Facility at the Japan Atomic Energy Research Institute, Naka, Japan in September-October 2001. Some measurements were performed also on the CSMC to study effects of electromagnetic and cooldown cycles. The TF Insert coil was charged successfully, without training, in the background field of the CSMC to the design current of 46 kA at 13 T peak field. The TF Insert met or exceeded all design objectives, however some interesting results require thorough analyses. This paper presents the overview of main results of the testing-magnet critical parameters, joint performance, effect of cycles on performance, quench and some results of the post-test analysis.

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Abstract-In this paper we report the main test results obtained on the Poloidal Field Conductor Insert coil (PFI) for the International Thermonuclear Experimental Reactor (ITER), built jointly by the EU and RF ITER parties, recently installed and tested in the CS Model Coil facility, at JAEA-Naka. During the test we (a) verified the DC and AC operating margin of the NbTi Cable-in-Conduit Conductor in conditions representative of the operation of the ITER PF coils, (b) measured the intermediate conductor joint resistance, margin and loss, and (c) measured the AC loss of the conductor and its changes once subjected to a significant number of Lorentz force cycles. We compare the results obtained to expectations from strand and cable characterization, which were studied extensively earlier. We finally discuss the implications for the ITER PF system. Index Terms-Cable-in-conduit conductors, fusion reactors, Nb-Ti superconducting material, superconducting magnets. I. BACKGROUND ON ITER PF CONDUCTORST HE ITER Poloidal Field (PF) conductors have undergone a significant evolution in the past years. In the original ITER design (2001) the Cable-in-Conduit Conductors (CICCs) were optimized to match the current/field levels in each of the six PF coils [1]. Following recent design reviews, a number of modifications have been introduced [2], leading to the conductor designs detailed in Table I, for the envelope of operating conditions in the PF Coils reported in Table II. The main change with respect to the original design is a reduction in the Cu:nonCu ratio of the low field conductors (PF2 to PF5), implying that the Stekly condition of cryogenic stability [3] is no longer respected. Experiments on subsize conductors [4] have suggested that in the planned regime of operation, and for the expected perturbation spectrum, full cryostability (i.e. a copper fraction corresponding to the Stekly limit) is excessive. In fact, for the conditions considered, it is more convenient to design the conductor for larger temperature margin, increasing the fraction of Nb-Ti, while maintaining the copper fraction to the strict minimum demanded for protection. To achieve the operating requirements of Table II, two main conditions must be met. Firstly, the cable performance must be close to the sum of the projected performance of the individual strands, without the occurrence of the premature quenches often seen in large size Nb-Ti conductors and attributed to current non-uniformity [4], [5] (see also later discussion). In practice, we quantify this first condition using the temperature margin above the operating temperature. The design value of the temperature margin is 1.5 K, with a maximum uncertainty of 0.5 K, which results in a minimum acceptable margin of 1 K.Secondly, all AC loss sources in the cable must be controlled, so to limit the temperature increase due to the heating due to the pulsed operation. In particular, the product of the cable demagnetization factor and coupling time constant, , proportional to coupling loss, must be smalle...