International Thermonuclear Experimental Reactor Central Solenoid Model Coil Research Articles

Preparation of the ITER Poloidal Field Conductor Insert (PFCI) Test

The Poloidal Field Conductor Insert (PFCI) of the International Thermonuclear Experimental Reactor (ITER) has been designed in the EU and is being manufactured at Tesla Engineering, UK, in the frame of a Task Agreement with the ITER International Team. Completion of the PFCI is expected at the beginning of 2005. Then, the coil shall be shipped to JAERI Naka, Japan, and inserted into the bore of the ITER Central Solenoid Model Coil, where it should be tested in 2005 to 2006. The PFCI consists of a NbTi dual-channel conductor, almost identical to the ITER PF1 and PF6 design, /spl sim/45 m long, with a 50 mm thick square stainless steel jacket, wound in a single-layer solenoid. It should carry up to 50 kA in a field of /spl sim/6 T, and it will be cooled by supercritical He at /spl sim/4.5 K and /spl sim/0.6 MPa. An intermediate joint, representative of the ITER PF joints and located at relatively high field, will be an important new item in the test configuration with respect to the previous ITER Insert Coils. The PFCI will be fully instrumented with inductive and resistive heaters, as well as with voltage taps, Hall probes, pick-up coils, temperature sensors, pressure gauges, strain and displacement sensors. The test program will be aimed at DC and pulsed performance assessment of conductor and intermediate joint, AC loss measurement, stability and quench propagation, thermal-hydraulic characterization. Here we give an overview of the preparatory work toward the test, including a review of the coil manufacturing and of the available instrumentation, a discussion of the most likely test program items, and a presentation of the supporting modeling and characterization work performed so far.

Modeling of thermal-hydraulic effects of AC losses in the iter central solenoid insert coil using the M&M code

During 2000, AC losses and the effects of possible ramp-rate limitation (RRL) were investigated on the International Thermonuclear Experimental Reactor (ITER) Central Solenoid Insert Coil (CSIC), at JAERI Naka, Japan. The CSIC was mounted inside the bore of the ITER Central Solenoid Model Coil (CSMC), at the maximum field of about 13 T and experiencing the largest magnetic field variations. The thermal-hydraulic response of the coil to different transport current scenarios was assessed by measuring the temperature increase and pressurization of the supercritical helium (SHe) coolant, together with the evolution of the mass-flow rate. Here we implement in the M&M code a detailed general model of AC losses, which is being validated for the first time. The resulting tool is then applied to the analysis of two CSIC tests, with different ramp-up of the transport current followed by the same dump, and used to qualitatively assess the major thermal-hydraulic effects of AC losses in the coil.

Analysis and interpretation of the full set (2000–2002) of TCS tests in conductor 1A of the ITER Central Solenoid Model Coil

The full set of T CS measurements performed during 2000–2002 on conductor 1A of the ITER Central Solenoid Model Coil (CSMC) of the International Thermonuclear Experimental Reactor (ITER) is analysed with the extensively validated M&M code. Under the assumptions of uniform strand properties and uniform current distribution among strands, the performance of this “average” strand in the cable is deduced from the best fit of the measured voltage–inlet temperature characteristics. Two fitting parameters are chosen: an ad-hoc additional contribution ‘ ε extra’ to the longitudinal strain of the average strand in the cable, and the average-strand (or cable “effective”) index ‘ n’ of the electric field–current density characteristic. It is shown that the average strand in the CSMC performed less well than expected from the strand database––an increasingly negative ε extra being needed to reproduce the coil behaviour at increasing transport currents, and that the cable effective n is clearly below the measured n of the strand. It is argued that this average-strand performance reduction in the CSMC is most likely related to mechanical load effects.

First test results for the ITER central solenoid model coil

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.

ＩＴＥＲ‐ＣＳモデル・コイルとＣＳインサート・コイルの実験結果 ＣＳモデル・コイルとＣＳインサート・コイルの初期冷凍・定常熱負荷・昇温特性

In International Thermonuclear Experimental Reactor (ITER) Engineering Design Activities, the Central Solenoid Model Coil (CSMC) and the CS Insert Coil (CSIC) have been fabricated, and installed into the ITER CSMC test facility at Naka Fusion Establishment, Japan Atomic Energy Research Institute (JAERI). The CSMC and CSIC conductors are applied to the cable-in-conduit conductors cooled by forced-flow supercritical helium (SHe) at 4.5K. There are 48 parallel cooling channels for the CSMC, the CSIC and the structures. In the CSMC experiment, the cool-down and the warm-up were finished successfully. The cool-down time was within 480 hours. The steady head load without coil current was measured and evaluated.

Mechanical properties of the ITER central solenoid model coil insulation under static and dynamic load after reactor irradiation

The candidate insulation system for the central solenoid (CS) model coil of the International Thermonuclear Experimental Reactor (ITER) is CTD-112P, an epoxy with a two-dimensionally woven S-glass-fiber reinforcement. Because of the pulsed operation of ITER, both the static and the fatigue behavior of the material have to be assessed under the actual operating conditions, including the appropriate radiation environment at the magnet location. To obtain information on the radiation-induced material degradation, the material was irradiated at ambient and low temperatures to neutron fluences of 5 × 10 22 m −2 ( E > 0.1 MeV). After measurements of swelling and weight loss, all experiments were carried out at 77 K. Half of the 5 K irradiated samples were subjected to a warm-up cycle to room temperature before testing. Tensile and short-beam-shear tests characterize the material in tension and interlaminar shear. Stress–lifetime diagrams of tensile and double-lap-shear specimens were assessed under tension–tension fatigue load up to 10 6 cycles.

Predicted thermohydraulic performance of the ITER Central Solenoid Model Coil conductors

The Central Solenoid Model Coil (CSMC) of the International Thermonuclear Experimental Reactor (ITER) will be installed and tested in 1999-2000 in a test facility at the Japan Atomic Energy Research Institute (JAERI) in Japan. The expected cable performance is predicted based on the strand data on RRR, n-value, critical current density, hysteresis loss measured for each production batch, as well as AC loss characterization on full size conductor samples. Simulation of the high field layer conductor performance has been performed for operation at 4.5 K by a 1-D thermohydraulic flow analysis code which also includes elements of the cryoplant facility. The simulation results show that the temperature of the conductor after the operation rises up to 6.0 K from the initial temperature of 4.5 K due to ac losses of the conductor, and the pressure rises up to 0.64 MPa from the initial pressure of 0.3 MPa. Thus, the simulation can correlate the strand and conductor data base with the thermohydraulic behavior of the Model Coil, and the experiments in turn can validate the simulation.

Electrical insulation systems for the ITER Central Solenoid Model Coil

Insulation systems developed by Composite Technology Development, Inc., have been fabricated specifically for use in the US section of the International Thermonuclear Experimental Reactor (ITER) Central Solenoid (CS) Model Coil. The insulation system used in these fusion devices must be capable of withstanding significant mechanical loads at cryogenic temperatures and enduring high doses of radiation, all without compromising their electrical insulation properties. Two materials that have seen extensive use in the CS Model Coil are the CTD-112P tetrafunctional epoxy pre-preg system and the CTD-618 epoxy system. Fabrication methods for these materials were developed and implemented specifically for use on the model cell. The CTD-112P system has been successfully used as the primary turn insulation, and as such, has been tested extensively for its mechanical and electrical properties, before and after irradiation. The CTD-618, a room-temperature cure system, has been specified as the insulation system for repair of damaged turn insulation and also as the insulation for the noncritical areas of the tension links and terminations. The development of these materials and the method of application for each of these materials as they apply to the model coil are examined. Pertinent mechanical and electrical properties are also analyzed.

Jacketing of the Superconducting Cables for the ITER Central Solenoid Model Coil (CSMC): Lessons Learned, Solved and Open Issues

In the framework of the activities for the International Thermonuclear Experimental Reactor (ITER) Ansaldo was given the order for the industralisation of the jacketing of the superconductors for the Central Solenoid Model Coil (CSMC). This task was successfully accomplished after solution of a number of problems, mainly in the area of the weldment of Incoloy 908.

ＳＭＥＳモデルコイル 初期冷凍・熱特性

A model coil of a superconducting magnetic energy storage (SMES) device, which is a forced-cooled Nb-Ti coil, has been fabricated and a performance test at cryogenic temperatures has been carried out. The SMES model coil is composed of 4 dual pancakes and its total weight is 4.5t. The applied conductors are cable-in-conduit conductors cooled by supercritical helium (SHe) at 4.5K and 0.7MPa. SHe is supplied to the SMES model coil and the structure by a reciprocating bellows pump. The test facility is located at the International Thermonuclear Experimental Reactor (ITER) common test facility, was constructed for the testing of an ITER central solenoid model coil. In the experiments, cool-down was finished within 10 days under controlled temperature differences in the SMES model coil. During cool-down and 4.5K operation, pressure drop characteristics of the conductor were measured and the friction factor estimated. The pressure drop characteristics of the SMES model coil were in good agreement with those of the previous cable-in-conduit conductor. During static operation without current, the heat load and refrigerator operation conditions were measured. The heat load of the SMES model coil is 7.5W, which is within the expected value.

Design of the ITER central solenoid (CS) model coil

The Japan Atomic Energy Research Institute (JAERI) has been designing and developing the central solenoid (CS) model coil for International Thermonuclear Experimental Reactor (ITER). The CS model coil is composed of two modules and they will be manufactured by a United States Home Team and a Japanese Home Team. The coil winding method is two-conductors-in-hand and the coil, which has an inner diameter of 1.6 m and an outer diameter of 3.6 m, produces a maximum field of 13 T. The coil will be tested in the JAERI facility. This paper describes the coil design, manufacturing procedure, stress analysis on the winding pack and AC loss analysis.

Results of simulations of stability and quench behavior for pulsed tests of the ENEA 12T Nb3Sn CICC coil

In the ENEA Frascati Laboratory a facility is being assembled to test the “ENEA Nb3Sn CICC coil” in pulsed regimes. The characteristics of the coil (dimensions, cable-in-conduit conductor, strand designed for use in variable field) are such to make these tests of primary importance to predict the behaviour of the ITER (International Thermo-nuclear Experimental Reactor) Central Solenoid Model Coil, which will be built in the next two years. In particular, the stability and quench behaviour of the coil will be tested and compared to the predictions of the thermo-hydraulic analysis code SARUMAN. Other important parameters will be the ramp rate limination, the limiting current and the conductor losses. Several testing scenarios (ramp up and discharge) are described and the present status of testing programme definition is given, together with the associated analyses.