E. Tada

Experimental results of the Nb3Sn demo poloidal coil (DPC-EX)☆

Abstract The aim of the development of DPC-EX is to demonstrate the applicability of an Nb 3 Sn conductor to pulsed coils for tokamak fusion machines. The DPC-EX, whose inner diameter is 1 m, consists of two double pancakes fabricated by a react-and-wind technique. The conductor is a forced cooled, flat, cable-in-conduit conductor. The DPC-EX has been installed between two Nb-Ti demo poloidal coils (DPC-U1 and U2). In the series operating mode (DPC-EX, DPC-U1 and U2), DPC-EX was ramped up to 17 kA in 1 s and ramped down to zero in 1 s after a flat top time for 1 s, without normal transition. The maximum magnetic field and the maximum pulse field were 6.7 T and 6.7 T s −1 , respectively. The current density in the winding was 37.2 A mm −2 at an operating current of 17 kA. The ratio of a.c. losses to the stored energy was ≈ 0.14% during pulsed operation. After more than 50 cycles of pulsed operation, no damage could be found in the DPC-EX. The stability test indicates that DPC-EX has a high stability margin. These results demonstrate the possibility of high field (12–14 T) poloidal coils, as required in the FER and ITER.

Mechanical Evaluation of Nitrogen-Strengthened Stainless Steels at 4 K

Because the structural materials of large superconducting magnets for fusion machines suffer high magnetic stress at 4 K, the knowledge of their mechanical properties at cryogenic temperature is highly desirable. Austenitic stainless steels have good properties for use in superconducting fusion magnets, but their relatively low yield stress at 4 K is serious drawback. Though some studies1,2 show this shortcoming can be overcome by adding nitrogen and carbon, there is as yet no reliable data base regarding the structural properties of the improved steels. During the design of the Japanese LCT coil, it became clear that the yield stress of the structural steel must be greater than 700 MPa at 4 K. For this reason, JAERI began to evaluate the mechanical properties of nitrogen-strengthened stainless steels at 4 K.

Development of Cryogenic Structural Materials for Tokamak Reactor

The structural materials of the toroidal field (TF) coils for fusion reactor are required to have higher strength and fracture toughness than those of ordinary commercial products, mainly because of large electromagnetic forces. During the preliminary design of the Japanese Large Coil Task (LCT) coil at the Japan Atomic Energy Research Institute (JAERI), ordinary 304L stainless steel was selected as the structural material for the LCT coil case. However, in the detailed design considering the winding rigidity and the slipping effect between the winding and the coil case, the stress intensity on the coillcase was increased more than that of the preliminary analysis. Finally, yield strength more than 700 MFa at 4 K was specified for the LCT structural material on the basis of many verification tests and analyses. Accordingly, ordinary 304L stainless steel could not be employed because of its relatively low yield strength at 4 K.

Domestic test result of the Japanese LCT coil

Japan Atomic Energy Research Institute (JAERI) has been preparing one D shape superconducting coil for the Large Coil Task. This paper describes mainly the results on domestic test which has been successfully carried out with a single test condition in JAERI this year. The main results, which were obtained during the test, are cool-down and warm-up characteristics, superconducting recovery characteristics, discharge characteristics, strain and displacement measurements, and heat load measurements. Before describing the results, the Japanese coil design parameters and the Superconducting Engineering Test Facility for the domestic test are shown in this paper.

Design and Development of the ITER Vacuum Vessel

Abstract In ITER, the vacuum vessel (VV) is designed to be a water cooled, double-walled toroidal structure made of 316LN stainless steel with a D-shaped cross section approximately 9 m wide and 15 m high. The design work which began at the beginning of the ITER-EDA is nearing completion by resolving the technical issues. In parallel with the design activities, the R&D program, Full-scale VV Sector Model Project, was initiated in 1995 to resolve the design and fabrication issues. The full-scale sector model corresponds to an 18° sector (9° sub-sector×2) and is being fabricated on schedule. To date, 60% of the fabrication had been completed. The fabrication of full-scale model including sector-to-sector connection will be completed by the end of 1997 and performance tests are scheduled until the end of ITER-EDA. This paper describes the latest status of the ITER VV design and the Full-scale Sector Model Project.

Recent progress in the demo poloidal coil program

The fabrication of two 30-kA NbTi pulsed coils and one 10-kA Nb/sub 3/Sn coil with a total stored energy of 40 MJ is in progress as a part of the DPC (Demonstration Poloidal Coil) Program. All the Nb-Ti superconducting strands have been fabricated, and their loss time-constant has been measured at 0.32 ms at 7 T, which is well below the initial target of less than 1 ms. A novel winding technique is described for the large current conductor which provides good mechanical contact. >

Development work for the Japanese LCT coil and its design and construction

Abstract This Paper describes the design, verification tests, and construction of the Japanese test coil for the Large Coil Task (LCT) under the auspices of IEA. The Japanese LCT coil was designed to be a pool-cooled, Nb-Ti, fully-stabilized coil with an operating current of 10 220 A at 8 T. Through research and development of the Japanese LCT coil, new advances in superconducting coil technology were acquired, such as mechanically and chemically treated conductor surfaces that have high heat transfer (about four times greater than usual). During the construction of the coil, new fabrication techniques were developed to wind the large-current conductor into a mechanically rigid coil and to construct a totally stable large coil. These advantages made it possible to construct the Japanese LCT coil which was completed in the spring of 1982.

Test results of the DPC-TJ: thermal and hydraulic performance☆

Abstract The thermal and hydraulic test results of the DPC-TJ coil, a 24 kA-40 A mm −2 forced-cooled Nb 3 Sn coil, are presented in this paper. The DPC-TJ coil was installed and tested between the DPC-U1 and U2 coils. The weight of the DPC-TJ coil is 2.8 ton and the total cool-down weight of the coil system is 23 ton. It took ≈ 180 h to cool the coils from room temperature to 20 K. The DPC-TJ coils heat load was 20 W at zero transport current and 40 W at 24 kA, the rated charging state. The pressure drop in the DPC-TJ coil was measured and the obtained value agreed well with that determined by the empirical formula used for the design of the DPC-TJ coil. A.c. losses and inductive losses due to heat input cause a rapid decrease in the helium flow rate at the coil inlet position. This phenomenon was analysed and one of the design standards was derived.

Experimental results of the Nb/sub 3/Sn demo poloidal coil (DPC-EX)

In order to demonstrate the applicability of a Nb/sub 3/Sn conductor to pulsed poloidal coils for Tokamak fusion machines, the Nb/sub 3/Sn Demo Poloidal Coil (DPC-EX) has been fabricated and tested. DPC-EX, whose inner diameter is 1 m, consists of two double pancakes made by a react-and-wind technique. The coil has a flat cable-in-conduit conductor cooled by forced-flow helium. DPC-EX has been set up in the Demo Poloidal Coil Test Facility (DPCF). In the operation of the coils in series, DPC-EX has been ramped up to 17 kA in 1 s. The magnetic field at this point was 6.7 T, and pulsed operation at 6.7 T/s was thus demonstrated. An average current density of 37.2 A/mm/sup 2/ in the winding was achieved in this operation, and the AC loss of the coil was confirmed to be quite small.

Construction and operation of the cluster test facility

The 20MJ superconducting cluster test facility, which has been under construction for the last two years at Japan Atomic Energy Research Institute (JAERI), was successfully operated in August 1980. Operating design values were achieved without any premature quenching. The fabrication and operation of the facility was the first experience with a superconducting toroidal coil in Japan. The succesful operation of the facility system makes it possible to test a 10T Nb 3 Sn magnet as a Test Module Coil which will be placed in the facility next year. This paper outlines briefly the facility system and describes mainly thermal results and stress results.