T. Hiyama

Design and fabrication of superconducting cables for ITER central solenoid model coil

The Nb/sub 3/Sn cable is being fabricated for the central solenoid (CS) model coil under the ITER Engineering-Design Activity. The cable consists of about 1000 strands whose diameter is 0.81 mm. The design current is 48 kA at a magnetic field of 13 T. The 0.6-GJ CS model coil is operated in a pulse mode (0.5 T/s). The first trial fabrication of a 100-m dummy cable and a 20-m superconducting cable was completed successfully. The second trial fabrication of a 1000-m dummy cable was performed to establish the stable manufacturing procedure in January, 1995. The authors measured the AC losses of the full-sized conductor and could determine the cable coupling time constant. They analyzed the heat generation of the CS model coil and calculated the temperature rise of the cable for the model coil.

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.

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. >

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.

Testing of the pool-boiling cooled Japanese LCT coil in the international fusion superconducting magnet test facility

The Japanese LCT coil has successfully achieved the design field of 8 T in the toroidal array of the six LCT coils. In March, 1986, tests of the LCT coils were started following the cooldown of the whole facility. In the single-coil test and design-point test, in which the maximum field was 6.4 T and 8.1 T, respectively, the coil was charged, discharged, and dumped under different conditions. A recovery test and a simulated nuclear heating test were performed to investigate the stability of the coil. Measurements of displacement under the several conditions of electromagnetic load were analyzed to explain the mechanical behavior of the coil. Results obtained in the tests are reported in this paper.

AC loss results of the Nb/sub 3/Sn Demo Poloidal Coil (DPC-EX)

AC losses in the DPC-EX, whose purpose is to demonstrate the applicability of Nb/sub 3/Sn conductor to poloidal coils for Tokamak fusion machines, were measured, and the results were analyzed. The coil was fabricated with a Nb/sub 3/Sn cable-in-conduit conductor and has a 1 m inner diameter. The AC loss was 9.4 kW for the pulse operation of a 0-17 kA-0 cycle with a ramp time of 1 s, corresponding to a 6.7-T maximum field and a 6.7-T/s pulse rate. It was 0.14% of the magnetic stored energy. The effective Nb/sub 3/Sn filament diameter was estimated as 32 mu m and the effective time constant of the coupling current within the strand was calculated as 2 ms on the assumption that coupling current between strands is negligible. The limitation of coil current due to AC losses is discussed. >

Experiment of 10-T, 60-cm-bore Nb 3 Sn test module coil (TMC-1) for the cluster test program

A 60-cm-bore coil wound with a reacted multifilamentary Nb 3 Sn conductor, named as TMC-1, was constructed. A magnetic field of 10.2 T was successfully generated at a current of 6,056 A with a back-up field of 3.3 T from the cluster test coil. The total stored energy was 39 MJ. The strain of the Nb 3 Sn conductor was 0.67 % including a bending strain of 0.54 % during winding. Moreover, a 30-cm-length normal zone, nucleated by heater-input technique in the innermost turn, was recovered to superconducting state at 10 T. This means that a heat flux of conductor cooling surface is more than 1.08 W/cm2. For the manual dump with a decay time of 14 second(B=0.48 T/sec.), the TMC-1 was stable without any damage. The TMC-1, which is pool-cooled at 4.2 K, is constructed as a step of the development of high field toroidal coil in a tokamak fusion machine. From thease results, it is demonstrated that multifilamentary Nb 3 Sn conductor is applicable to large-current and large-size coil.

Cryogenic system development and helium behavior study for forced-flow superconducting coils

In Japan Atomic Energy Research Institute (JAERI), cryogenic technology development is propelled to aim at realization of superconducting coil system for fusion experimental reactor (FER). For this purpose, forced-cooling technology which is one of attractive cooling methods and is expected to use for one of large superconducting coils for fusion is being investigated according to the cryogenic technology development program shown in Fig. 1. JAERI has already constructed and tested three forced-flow generating facilities which are named as forced flow generator (FFG), segment test facility (STF) and forced flow test facility (FFTF). The forced flow generator (FFG) which can provide supercritical helium up to 3 g/s with 8 atm and 4.5 K was firstly fabricated for fundamental investigation of forced-cooling coils and coolant. As the second step, STF and FFTF were constructed in order to investigate coolant, supercritical helium, control technique combined with the helium liquefier/refrigerator. They are designed to have the capacities of flow rate up to 20 g/s and 60 g/s with 15 atm and 4.4 K by adding supercritical heat exchanger to the existing 350-l/h and 1.2-kW helium cryogenic system. Using these facilities, several forced-cooled superconducting coils with cable-in-conduit conductor were tested and the stability characteristics and supercritical helium behavior in the conductor were measured. This paper describes design concept and tested performances for the forced flow facilities, and pressure rise Characteristics of supercritical helium in cable-in-conduit conductors.

Boiling helium heat transfer characteristics in narrow cooling channel

For design of a stable superconducting magnet with pool cooling method, the size of the cooling channel is one of the important factors to determine its overall current density. We measured the steady state boiling helium heat transfer characteristics in a 600 mm long vertical cooling channel. The result shows that heat transfer characteristics in film boiling region do not become worse so far as the channel is not extremely narrow and rather becomes better adding some amount of bubbles in the channel, whereas that in nucleate boiling region becomes worse monotoneously and rapidly as the thickness of the channel decreases. From these results, it proves to be much better to choose the superconductor whose critical temperature is high to get the high current density stable magnets and also proves to be able to design stable pool-boiling magnets as large as toroidal coils for Tokamak machines of the next generation.