T. Ando

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.

Pulsed field loss characteristics of the Japanese test coil for the large coil task

This paper describes some results of an analytical calculation and the measurements of a pulsed field loss on the Japanese LCT conductor. The conductor, whose aspect ratio is around 2.5, is wound edge wise to reduce the pulsed field loss. The Japanese LCT coil, with the conductor thus designed, was evaluated to have a pulsed field loss of about 19W. This includes the loss of the helium vessel under the normal LCT test with a pulsed field of 0 to 0.14T. In addition, the loss measurement has been carried out up to 25 T/s. Measured results on the relation between the loss of a single strand and that of a whole strand pack well supports the validity of the loss analysis which includes a newly arranged set of analytical equations.

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.

Development of a forced-cooled superconducting coil with high average current density (DPC-TJ)

A forced-cooled superconducting coil (DPC-TJ) using a double-walled cable-in-conduit, the so-called Performed Armor CICC, has been developed. The DPC-TJ is an advanced coil with an average current density of 40 A/mm/sup 2/, operating current of 24 kA, and field of 12 T. The DPC-TJ coil has several advantages such as mechanical rigidity, negligible degradation of critical current, sound electrical insulation, and excellent winding tolerance compared with a conventional forced-cooled coil. A superconducting strand was developed with the aim of improving the critical current density and resistivity. The resulting Nb-tube-processed (NbTi)/sub 3/Sn strand fulfils the requirement that the critical current density be more than 600 A/mm/sup 2/ at 12 T and the RRR (residual resistivity ratio) be higher than 50. >

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.

Development of 12 T-10 Al-stabilized Nb 3 Sn conductor for TMC-II

The Al-stabilized Nb 3 Sn strand has been successfully fabricated and J c value of 400 A/mm2at 12 T is obtained on condition that Nb 3 Sn is reacted at 625°C for 200 hr. Overall residual resistivity of this strand is lowered to 77 % of that of the Cu-stabilized Nb 3 Sn strand (TMC-I)1,2at 12 T by Al-stabilizer. The 12 T-10 kA cable-in-conduit conductor, fabricated by this strands, is charged up to 20 kA at 8.7 T without appearance of normal zone. From these experimental results, this conductor satisfies almost the specification of the present target conductor (TMC-II)3.

Effect of the thermal barrier on the stability of cable-in-conduit conductors

As the enhancement of stability margin of cable-in-conduit conductors, the effect of thermal barrier provided inside the surface of strand has been investigated. Firstly, by a simple model calculation, it is shown that the energy needed to raise the temperature of a strand as the simulation of the disturbance due to surface friction, depends strongly on the pulse duration and thermal conductivity of surface material. Next, as based on this consideration, strands with the thermal barrier inside its surface have been made and the stability test has been carried out. From the result, it was shown that the stability margin of a cable-in-conduit conductor consisting of strands with the thermal barrier can be calculated with the enthalpy of whole helium whithin the conduit even near the critical current.

Development of a 30-kA cable-in-conduit conductor for pulsed poloidal coils

This paper describes designed parameters of a 30-kA cable-in-conduit conductor (JF-30), and the test results of stability margin measured by using a triplex in a conduit. Cross sectional size of JF-30 is 35mm × 35mm and 567 NbTi-Cu-CuNi strands are in a stainless steel conduit whose thickness is 2 mm. Void fraction is 33 % and the designed stability margin is 270 mJ/cc at 5 atm and 7 T. Stability test by a triplex showed a favorable margin, a few hundreds of mJ at 7 T even without helium flow. In addition, the stability was strongly increased when helium flow up to 0.2 g/s was applied. At around 3 atm, we found that the stability margin was mere than 2 J/cc which exceeded the present heater capacity. This resulted in an extension of current range, in which the sample is stable, up to 150 to 200 % when compared to the case without helium flow.

Test results and perspectives of the cluster test program

The Cluster Test Facility has been constructed and successfully tested at the Japan Atomic Energy Research Institute (JAERI) for the development of superconducting tokamak toroidal coils. The Cluster Test Coils, constructed in 1980, are 20MJ, NbTi, pool-cooled coils. During the first test operation, the two coils were charged successfully to the rated current value of 2,145A without quenching. The detailed data obtained during the cool down was significant. This information pertains mainly to the efficiency of the cooling surface and thermal properties of the coil legs composed of stainless steel and glass epoxy. Results of strain measurements during the charging tests indicated a serious problem of analytical calculations. This problem arises primarily from the difficulty of defining the mechanical properties of winding-structure interface. The first test module coil is a 10T, Nb 3 Sn coil. This coil is now under construction and will be tested in the Cluster Test Facility in 1982.