M. Oshikiri

Progress of the ITER central solenoid model coil programme

The worlds 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.

Test of the ITER TF insert and Central Solenoid Model Coil

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.

First test results for the ITER central solenoid model coil

Abstract The largest pulsed superconducting coils ever built, the Central Solenoid (CS) Model Coil and Central Solenoid Insert Coil were successfully developed and tested by international collaboration under the R&D activity of the International Thermonuclear Experimental Reactor (ITER), demonstrating and validating the engineering design criteria of the ITER Central Solenoid coil. The typical achievement is to charge the coil up to the operation current of 46 kA, and the maximum magnetic field to 13 T with a swift rump rate of 0.6 T/s without quench. The typical stored energy of the coil reached during the tests was 640 MJ that is 21 times larger than any other superconducting pulsed coils ever built. The test have shown that the high current cable in conduit conductor technology is indeed applicable to the ITER coils and could accomplish all the requirements of current sharing temperature, AC losses, ramp rate limitation, quench behavior and 10 000-cycle operation.

Development of 46-kA Nb/sub 3/Sn conductor joint for ITER Model Coils

The conductor joint is one of the key technologies for superconducting coils. A butt type joint has been successfully developed for the ITER magnets. The 46 kA Nb/sub 3/Sn conductors are connected by the diffusion bonding technique in vacuum, after the reaction of Nb/sub 3/Sn. The advantage of this joint is low losses against pulse field, because the compacted part is very small compared with other types of joint. 15 butt joints have already been fabricated in the ITER CS Model Coil. According to the test results of the full-size conductor samples, these butt joints will be operated stably in the pulse operation, because the temperature increase due to ac losses and Joule heating by joint resistance is very small and the joint has a sufficiently high temperature margin.

Test of the NbAl insert and ITER central solenoid model coil

The Central Solenoid Model Coil (CSMC) was designed and built by an ITER collaboration in 1993-2001. Three heavily instrumented Inserts have been also built for testing in the background field of the CSMC. The Nb/sub 3/Al Insert was designed and built by Japan to explore the feasibility of an alternative to Nb/sub 3/Sn superconductor for fusion magnets. The Nb/sub 3/Al Insert coil was tested in the CSMC Test Facility at the Japan Atomic Energy Research Institute, Naka, Japan in March-May 2002. It was the third Insert tested in this facility under this program. The Nb/sub 3/Al 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 and later was successfully charged up to 60 kA in 12.5 T field. This paper presents the test results overview.

The Second Test Results on the Nb3Sn Demo Poloidal Coil(DPC-EX)

The second test of the DPC-EX fabricated as one of developments of superconducting poloidal coils for fusion machines, has been carried out. The DPC-EX is a winding inner diameter of 1m, wound with a Nb3Sn cable-in-conduit conductor. In the test, the generation of 7.1 T in 0.5 s was achieved with a high cable space current density of 117 A/mm2. Measurement of ramp rate limitation was carried out and a simple formula was presented for analysis of the results. And it was found that no changes in Tcs and ac losses were appeared before and after 350 pulsed operations. These results show that the reliability of Nb3Sn conductors for large pulse coils was demonstrated.

Performance of Japanese Conductors for ITER Toroidal Field Coils

The cable-in-conduit conductors for the ITER TF coils are fabricated using the latest high performance strands. The strands made by bronze and internal-tin methods are used for the conductors with a void fraction of 29% and 33%, respectively. Superconducting performance of the conductors was measured at the operating condition of the TF coils. The measured current sharing temperatures Tcs are 6.3-6.6 K for the bronze and 5.6-6.1 K for the internal-tin. The Tcs of the conductor with void fraction of 29% is 0.1-0.3 K higher than the conductor with a void fraction of 33%. It is shown from the results that the strain on the cable is between 0.7% and 0.75% and the n-values are between 4 and 6, much smaller than the n-values of strands.

EXPERIMENTAL INVESTIGATION OF PRESSURE RISE OF QUENCHING CABLE-IN-CONDUIT SUPERCONDUCTOR

Pressure rise during the quench of cable-in-conduit conductors has been experimentally studied. The cable consists of eighteen NbTi/Cu strands and one stainless steel wire. Experiments were carried out for two cases: in the first case, an entire conductor length goes normal simultaneously by using a resistive heater installed inside the conduit, and in the second case, quench starts with local normal zone created by using an inductive heater located at the center of the conductor. From the former experiment, the pressure rise was formulated as a function of time. In the later experiment, the dependency of transport current, magnetic field and initial pressure on the pressure rise were investigated. The rate of the pressure rise is found out to increase in proportion to the transport current.

Downstream effect on stability in cable-in-conduit superconductor

Abstract Japan Atomic Energy Research Institute (JAERI) conducted stability measurements of a 15 kA superconducting test loop with a cable-in-conduit conductor cooled by forced-flow supercritical helium. In this stability experiment, a disturbance to initiate normalcy is applied by an inductive heater set on the conductor near the coolant inlet so as to partially heat it within 6 ms. We have found that stability of a cable-in-conduit conductor is dominated by two types of quench, that is, quench at the heated zone and quench at the downstream zone. The quench at the downstream zone is caused by hot helium flowing from the heated zone in the upstream zone and is serious for forced-flow cooling coils because the stability margin of the downstream zone is lower than that of the heated zone. The measured stability characteristics and the comparison with the analytical results are described in this paper.

Development of the New Cryogenic Structural Material for Fusion Experimental Reactor

The structural material of the superconducting coil for Fusion Experimental Reactor (FER) requires higher mechanical strength and fracture toughness than those of ordinary austenitic stainless steels such as 304LN, 316LN at liquid helium temperature.1 Japan Atomic Energy Research Institute (JAERI) determined a target on engineering properties of structural material in FER coil system.2,3 The target is as follows.