P.C. Michael

Test of the ITER central solenoid model coil and CS insert

The Central Solenoid Model Coil (CSMC) was designed and built from 1993 to 1999 by an ITER collaboration between the U.S. and Japan, with contributions from the European Union and the Russian Federation. The main goal of the project was to establish the superconducting magnet technology necessary for a large-scale fusion experimental reactor. Three heavily instrumented insert coils were built to cover a wide operational space for testing. The CS Insert, built by Japan, was tested in April-August of 2000. The TF Insert, built by Russian Federation, will be tested in the fall of 2001. The NbAl Insert, built by Japan, will be tested in 2002. The testing takes place in the CSMC Test Facility at the Japan Atomic Energy Research Institute, Naka, Japan. The CSMC was charged successfully without training to its design current of 46 kA to produce 13 T in the magnet bore. The stored energy at 46 kA was 640 MJ. This paper presents the main results of the CSMC and the CS Insert testing-magnet critical parameters, ac losses, joint performance, quench characteristics and some results of the post-test analysis.

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.

ITER CS model coil and CS insert test results

The inner and outer modules of the central solenoid model coil (CSMC) were built by US and Japanese home teams in collaboration with European and Russian teams to demonstrate the feasibility of a superconducting central solenoid for ITER and other large tokamak reactors. The CSMC mass is about 120 t; OD is about 3.6 m and the stored energy is 640 MJ at 36 kA and peak field of 13 T. Testing of the CSMC and the CS insert took place at Japan Atomic Energy Research Institute (JAERI) from mid March until mid August 2000. This paper presents the main results of the tests performed,.

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.

Ramp-rate limitation experiments using a hybrid superconducting cable

Abstract Ramp-rate limitation experiments were done in a new facility at the MIT (Massachusetts Institute of Technology) Plasma Fusion Center. The features of this new facility include (1) a superconducting pulse coil that can superimpose high ramp-down rates, up to 25 T s−1, (2 T in 80 ms) at a background field up to 5 T, (2) new power supplies that can supply high rates of dl dt and dB dt to the sample under test and (3) a forced-flow supercritical helium system for cooling CICCs (Cable-In-Conduit Conductors). This paper discusses the results of the ramp-rate limitation experiments on a 27-strand hybrid Nb3Sn cable. The cable was tested under field ramps of up to 2.5 T s−1 with various operating currents. It did not quench with dB dt , field and average strand currents that were simultaneously above the operating range of TPX-PF (Tokamak Physics Experiment Poloidal Field) coils. Further ramp-rate limitation experiments revealed that the tested 27-strand hybrid cable has very high transient stability at ramped fields, extending out to average strand currents that are nearly triple the TPX-PF operating current.

CURRENT LEAD OPTIMIZATION FOR CRYOGENIC OPERATION AT INTERMEDIATE TEMPERATURES

The refrigeration power for large current superconductor systems, such as for electrical power distribution, is dominated by current lead losses. The use of multiple cooling stages between room temperature and ∼70 K is investigated as means to decrease the refrigeration power. We show that it is possible to decrease the electrical power requirements for the refrigerator by about 1/3 through the use of two‐stages current leads; this computed power saving is based on a conservative estimate of refrigerator performance. Using data from real systems, that is, higher temperature refrigerators operating at higher fractions of their Carnot efficiencies we believe that the refrigerator electrical power requirement can actually be decreased by 1/2. Means have been investigated to optimize current lead performance at lower than maximum current operation. Adjustment of the cooling power of the intermediate temperature refrigerators achieves limited success in power consumption minimization. Other means to optimize the performance will be described. The implications of intermediate stages for stability of the current leads following a short overcurrent period will be described.

Design, fabrication and test of the react and wind, Nb3Sn, LDX floating coil

The Levitated Dipole Experiment (LDX) is an innovative approach to explore the magnetic confinement of fusion plasma. A superconducting solenoid (floating coil) is magnetically levitated for up to 8 hours in the center of a 5-meter diameter vacuum vessel. The floating coil maximum field is 5.3 T, and a react-and-wind Nb/sub 3/Sn conductor was selected to enable continued field production as the coil warms from 5 K during the experiment up to a final temperature of about 10 K. The coil is wound using an 18-strand Rutherford cable soldered into a half-hard copper channel, and is self protected during quench. The coil is insulated during winding and then vacuum impregnated with epoxy. The impregnated coil is tested with 2 kA operating current at 4.2 K, and then a single, low resistance joint is formed at the outer diameter of the coil before the coil is enclosed in its toroidal helium vessel. This paper presents details of the coil design and manufacturing procedures, with special attention to the techniques used to protect the coil from excessive strain damage throughout the manufacturing process.

Design and manufacture of the US-ITER pre prototype joint sample

The US-ITER pre prototype joint sample which has been fabricated at the MIT Plasma Fusion Center is the first attempt to fabricate an optimized full size joint which can be stably operated in ITER required AC background fields at reduced coupling losses. This paper presents an overview of the joints construction and fabrication, highlighting some of the procedural steps that have since been incorporated into fabrication of current terminations for the inner module ITER central solenoid (CS) model coil.

TPX superconducting tokamak magnet system 1995 design and status overview

The TPX magnet preliminary design effort is summarized. Key results and accomplishments during preliminary design and supporting R&D are discussed, including conductor development, quench detection, TF and PF magnet design, conductor bending and forming, reaction heat treating, helium stubs, and winding pack insulation.

Reassessment of cryotribology theory

Abstract The interpretation of cryogenic temperature friction is enhanced by the concurrent evaluation of the low-temperature mechanical properties of sliding materials. Absolute sliding stability is greatly desired but difficult to attain in cryogenic systems. Experiments performed to investigate the material factors necessary for smooth motion indicate that interfacial shear creep plays a key role in sliding stabilization. Because viscous creep is typically not observed at cryogenic temperatures, we conclude that it is not possible to guarantee sliding stability based solely on material selection criteria.