Makoto Sugimoto

Impact of Cable Twist Pitch on

The performance of four Nb3Sn conductors for the ITER central solenoids was tested. The current sharing temperatures (Tcs) were measured over approximately 9000 electromagnetic cycles, including two or three thermal cycles between 4.2 K and room temperature. Tcs increased and became almost constant through the cycling. The gradient of the electric field against the temperature gradually decreased against cycling. The degradations caused by the electromagnetic force of the short twist pitch conductors were smaller than that of the original twist pitch conductor. The ac losses of short twist pitch conductors were several times higher than that of original twist pitch conductor. The dents and the removals of the Cr plating on the strands, which were formed during cabling, decreased the electric resistance between strands, which may cause the observed high ac loss. Inspection of the cable showed neither a clear bias of cable in the cross-sectional surface nor distorted strands in the lateral face. The high rigidity of the short twist pitch cable could prevent these plastic deformations, caused by the Lorentz force.

T_{cs}

Under the International Thermonuclear Experimental Reactor (ITER) project, the Japan Atomic Energy Agency (JAEA) is procuring all of the Nb3Sn conductors for the Central Solenoid (CS). The CS consists of six vertically stacked modules. The height and outer diameter of the CS are approximately 13 m and 4 m, respectively. The CS has a circular five stage cable. All of approximately 43 km of Nb3Sn CS cables will be manufactured in Japan. Before mass-production start, the jacketed cable conductors should be tested in the SULTAN facility in Switzerland to confirm their superconducting performance. The original cabling design had relatively long twist pitches and is referred to as the normal twist pitch (NTP) conductor. The NTP conductor test results revealed decreasing the current sharing temperature (Tcs) with increasing number of electro-magnetic (EM) load cycles. Therefore, a short twist pitch (STP) design was proposed and the STP conductors were also tested. The STP conductor results showed that the Tcs is stable during EM cyclic load tests. Because the conductors with STP have a smaller void fraction in the cable area than those with NTP, a higher compaction ratio during cabling is required and the possibility of damage on strands increases. The STP cable technology was developed in collaboration among Japanese cabling suppliers and JAEA. Several key technologies will be described in this paper.

-Degradation and AC Loss in

The Nb3A1 superconductor has been expected to the application under the high field, such as the one for International Thermo-nuclear Experimental Reactor (ITER), on the point of superior characteristics of Ic under stress. The present status of the development of Nb3A1 superconducting wires manufactured by the Jelly-Roll process is presented in this paper. The recent development has achieved the Nb3A1 superconductor stabilized by the copper matrix with a high critical current density and a low hysteresis loss. In the development, the microstructure of Nb-Al compound phases, each identified by EDS, are observed. The volume of each Nb-Al phase calculated by the Image Analysis System is discussed in terms of its dependence of Jc. With a non-copper critical current density of more than 550 A/mm2 at 12T, the wire features excellent high field characteristics, for example μ0Hc2 is 21.5T. Also, this wire has less degradation of Ic by strain compared to Nb3Sn. Hysteresis losses in the wire was measured by the magnetization method. For the field perpendicular to the wire length, it was shown that the effective filament diameter is almost equal to the actual filament diameter (21μm). On the other hand, the hysteresis loss for the parallel field was around one-fourth as large as that for the perpendicular one, corresponding to an effective filament diameter of approximately 8 µm. The 40kA-class cable-in-conduit conductor was fabricated to demonstrate its applicability to fusion magnets. The critical current test of the conductor proves that the capacity of 40kA is attained at 11.2T. The practical 10kA-–100m class long length cable-in-conduit conductor, of which conduit is Titanium, was fabricated first in the world.

\hbox{Nb}_{3}\hbox{Sn}

The central solenoid (CS) model coil-outer module being fabricated to demonstrate the justification of the CS design for the ITER, was almost completed except for epoxy impregnation to concrete whole layers. All the wound and heat treated layers have been assembled symmetrically with the insulation on the same axis, and for layer-to-layer joints the newly developed butt joint, has been installed.

Conductors for ITER Central Solenoids

Abstract The development work of the high-performance Nb3Sn superconducting strand for the Center Solenoid Coil, a high-field large pulse coil, of the International Thermonuclear Experimental Reactor (ITER) have been progressed by Japan Atomic Energy Research Institute with Japanese wire and cable companies. The main. required performances for the 0.8 mm-diameter chromium plated strand were 550 A/mm2 or higher as the critical current density at 12 T and 200 mJ/cm3 or lower as the hysteresis loss for ±3T.As the result of research and development work, the reliable manufacturing process of the high-performance strand was established, and the mass-production for the ITER Center Solenoid Model Coil was started.

Cabling Technology of

The inner and outer modules of the Central Solenoid Model Coil (CSMC) are mechanically compressed by a preload structure. The necessary preload is determined for the support of the modules terminal leads. The forces in the preload structure are monitored by strain gauge bridges mounted to each of the structures 16 inner and 16 outer tension rods during: (1) room temperature loading; (2) cool-down to the coils operating temperature; and (3) electromagnetic operation. The preload structure is described and the variations in the preload with each stage of coil operation are presented. Analytical models are presented, which use the observed variations in preload to deduce equivalent mechanical properties for the coil modules.

\hbox{Nb}_{3}\hbox{Sn}

Abstract The cable-in-conduit conductor (CICC) is applied for large devices, such as the fusion magnets, because it has a high mechanical and electrical performance potential. Also, CICC has advantages when applied for pulsed operation coils, such as the fusion magnets and superconducting magnetic energy storage coils, because of its thermal and hydraulic performance. Superconducting coils made from CICC have a cooling path and the coolant of the supercritical helium circulates through the inside of the CICC. When the coil has heat generation due to the AC losses, the CICC has the advantage of being able to remove heat due to the large wet perimeter. The inlet flow reduction is caused by the heat generation of the conductor. The inlet flow reduction provides a limit to the stable pulsed operation of the coil. The inlet flow reduction is measured and discussed for the CICC in this paper. The design criteria of the inlet flow reduction is introduced for the CICC. The zero flow temperature is defined to achieve this purpose. The initial flow rate can be estimated by this criteria, which is applicable for the forced flow conductor, as well as for CICC.

Conductor for ITER Central Solenoid

Abstract The flow reduction of forced flow superconducting coil with a cable in conduit (CIC) conductor has been studied for AC losses due to pulsed operation. In this paper, the flow reduction by rapid heating is described for forced flow superconducting coil with a CIC conductor. The phenomenon of flow reduction of the forced flow coil has been applied for coil quench detection and has been developed by Japan Atomic Energy Research Institute (JAERI). It is named the “fluid method” and essential technology for quench detection of large-scale forced flow superconducting coil as fusion magnets and superconducting magnetic energy storage (SMES) coil. In the fluid method, the inlet flow reduction is caused by Joule heating on the normal zone of superconducting coil. The fluid method has no electric noise in its detection. This is an advantage for pulsed operation in comparison with other electrical quench detection systems. On the other hand, there are no quantitative considerations between the inlet flow reduction and Joule heating of the coil inside in previous studies. The flow reduction for the quench detection has been determined by the operation experience of forced flow superconducting coil. The purpose of this paper is an estimation of the relation between the inlet flow reduction and Joule heating at coil quench. First, the inlet flow reduction was obtained by experiment in which the sample has an inductive heater for the quench emulation. Second, the evaluation model was proposed and its model showed good agreement between the inlet flow reduction and heat generation of the coil inside by rapid heating as coil quench.

Development of Nb3Al/Cu Multifilamentary Superconductors

Abstract In order to develop a superconducting magnet for a nuclear fusion reactor, it is necessary to assess the tribological properties of the structural materials. In this work the frictional properties and surface damage of new cryogenic structural steels, JN1 (25Cr-15Ni) and JN2 (15Cr-25Mn), were studied at 293 K in air, 77 K in liquid nitrogen and 4 K in liquid helium under fretting conditions. The fretting conditions were a normal load of 20 N, a slip amplitude of 100 μm, a frequency of 8.33 Hz and 5×10 4 cycles. The coefficient of friction decreased with a decrease in temperature for like metal combinations, and the values at 4 K were about 0.25 and 0.3 for JN1 and JN2 respectively, in spite of the occurrence of adhesive metal-to-metal contact. These results were quite different from those for copper and SUS316L. It seems that the low friction at 4 K results from the high hardness of JN1 or JN2 which prevents strong adhesion between surfaces in liquid helium. The coefficient of friction of metal fretted against GFRP was 0.35 at 4 K, but it was at the minimum of 0.25 at 77 K. The metallic surface suffered wear damage from glass fiber in glass fiber reinforced-epoxy.

Fabrication of ITER central solenoid model coil-outer module

Abstract Japan Atomic Energy Research Institute is working on the detail design of the helium refrigerator system (8 kW) in order to cool down Central Solenoid Model Coils (CS Model Coils), which will be fabricated and tested in International Thermonuclear Experimental Reactor (ITER) R&D project. The helium refrigerator system consists of supercritical helium cryogenic pump unit, helium refrigerator/liquefier unit, etc. The refrigerator system is designed, considering the various coil operation modes, refrigeration efficiency, economy, and research and development of the cryogenic components required for the ITER.