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The mechanism of the coupling loss is examined by measuring the trajectories of strands of sample CIC conductor. The studies have shown that probable loop lengths of long conductors are calculated by contact probabilities of two strands at each contactable point of two sub cables by newly developed function of twisting pitches.

The tests and analysis performed in Japan Atomic Energy Agency (JAEA) to solve the major technical issues in manufacturing the Toroidal Field coil winding pack (TFWP) are discussed. New impregnation method is under development for shortening the impregnation period.

The test results of the new generation of ITER TF conductor samples in SULTAN are discussed. The homogenization of performance toward the upper range has provided an effective margin for the operation of the ITER coil system.

The predictive analysis of the International Thermonuclear Experimental Reactor (ITER) Poloidal Field Conductor Insert (PFCI) test program is presented. PFCI is the first insert coil based on NbTi and is aimed at bridging the extrapolation gap between its full-size short sample and the ITER Poloidal Field (PF) coils.

A two-turn D-shaped test coil with a height of 2.0 m and a width of 1.8 m is fabricated by a react-and-wind method by using a 20 kA Nb3Al cable-in-conduit (CIC) conductor. The total strain of the D-shaped test coil is about 0.57 percent, the same as the triplex-CIC sample fabricated by a W&R method, indicating that bending did not influence the I(sub c).

Development of Nb3Sn cable-in-conduit (CIC) conductors with stainless steel jackets for the central solenoid and divertor coil of the superconducting tokamak called JT-60SC is described. The experimental results show that the characteristics of the full-size conductor can be predicted by measuring of J(sub c) - B - T characteristics of sub-size conductors.

Cable-in-conduit conductors (CICC) consisting of about 1000 Nb(sub 3)Sn strands with an outer diameter of bout 0.8 is designed for the TF and CS coils of the ITER. The test results indicate that both joint types are stable and could be used for applications to the full size ITER coils.

The ITER Central Solenoid, which was 12 m high and 4 m in diameter, consists of a stack of 6 electrically independent modules to allow control of plasma shape. The CS model coil and insert coil test results indicate that the conductor design must be modified to achieve an operation margin.

The world's largest high-field pulsed superconducting magnet, the CS Model Coil, is manufactured and tested by the International and Participant Teams for the verification of new design principles and fabrication concepts proposed for the ITER CS. Significant effects of large electromagnetic forces acting o the conductor are observed in the performances and the movement of cable inside the conductor is identified to be important.

Japan Atomic Research Institute has developed a low carbon and boron added 0.03C-22Mn-13Cr-9Ni-1Mo-0.24N-0.03B steel (JK2LB) for a conductor jacket of the central solenoid for the International Thermonuclear Experimental Reactor. The structural design and mechanical properties are discussed.

Tests of the International Thermonuclear Experimental Reactor (ITER) Toroidal Field (TF) Insert and Central Solenoid Model Coil (CSMC) are evaluated.

Performance of the NbAl Insert and the International Thermonuclear Experimental Reactor (ITER) Central Solenoid Model Coil (CSMC) are discussed.

An evaluation method for critical currents and current sharing temperatures is proposed, and it is applied to the ITER Center Solenoid (ITER-CS) Model Coil Insert.

Tests on the ITER project are described and discussed.

The International Thermonuclear Experimental Reactor's (ITER) Central Solenoid Model Coil (CSMC) project is described and discussed. All of the project's goals were successfully achieved.

The ITER-Central Solenoid (CS) model coil was fabricated and tested. Test results are presented and discussed.

First post-test analysis results of the Inner and Outer modules of the Central Solenoid Model Coil are reported.

Instrumentation in the vacuum chamber in the Central Solenoid Model Coil and the CS Insert coil operation is investigated.

The International Thermonuclear Experimental Reactor central solenoid model coil conductors were studied to determine predicted thermohydraulic performance.

Test results of the 460-kA Nb3Sn conductor joint for the ITER model coils are presented.

The completion of the outer module fabrication for the ITER CS model coil is discussed and the future connection to the inner module coil is considered.

The Nb3Al superconductor has been considered as one of the candidates for the Toroidal Field Coil superconductor in the International Thermonuclear Experimental Reactor (ITER) project. Testing of the Nb3Al insert will be done in the bore of the Central Solenoid Model Coil under ITER Engineering Design Activities. Using the jelly-roll process for the Nb3Al insert, a copper stabilized multifilamentary strand was developed. The strand had Jc above 600 A/mm2 at 12T under the bending strain of 0.4%.

A copper stabilizer was transferred from its location near the strand surface to the center of the strand's cross-section to enhance the residual resistivity degradation. The approach led to an enhancement of one-sixth of previous value to 1.6x10(super -10) ohmmeter in a laboratory scale strand. The strand exhibited a critical current density of 715 A/mm2 at 12 T. The approach has been applied in the International Thermonuclear Experimental Reactor program.

The Engineering Design Activity (EDA) of International Thermonuclear Experimental Reactor (ITER) has been carried out since 1992 among the four Parties which are European Union, Japan, Russian Federation, and United States. The design and the fabrication of the Center Solenoid (CS) model coil is under way in the ITER-EDA. The CS model coil consists of four parts, which are the outer module, the inner module, insert, and the supporting structure. The design of the CS insert coil is reported and discussed. The inner diameter of the CS insert coil is 1.5m. The conductor of the CS insert coil is the same as that of the ITER-CS coil. The CS insert coil will be installed inside the CS model coil. The conductor performance will be demonstrated on the maximum magnetic flux density of 13 T at the test facility of Japan Atomic Energy Research Institute (JAERI).

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 United States Home Team and 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.

The design and fabrication techniques for the ITER magnets are supported by the ITER Model Coil Program which focuses on component development and performance verification. The key component of the magnet system is the high current capacity (45 - 60 kA), [Nb.sub.3]Sn, cable-in-conduit conductor, which is now being fabricated under a staged development program. The integrated design features and performance of the 13 T Toroidal Field coils and Central Solenoid will be verified in the Model Coil Program. This paper describes the status of the structural and construction features of the model coils and the test facilities for the model coils at JAERI (Naka) and FZK (Karlsruhe).

Effect of strand diameter on the stability of the cable-in-conduit type superconductor was investigated experimentally. The cable-in-conduit type conductor is widely adopted in the design of large superconducting magnets and its properties have been studied. Present experiment was planned and performed to clarify the relation between the stability and the strand diameter or cooling perimeter. As a result of a series of present experiment, it was confirmed that the recovery capacity of the cable-in-conduit type superconductor from the normal transition was in proportion to the cooling perimeter of the cable.

The effect of cyclic pulsed operation on the coil performance was studied from the electrical and mechanical points of view in the second experiment of the [Nb.sub.3]Sn Demo Poloidal Coil (DPC-EX). Current sharing temperatures, ac losses, displacements, and strains were measured before and after cyclic pulsed operation, which consisted of about 350 cycles. No degradation of the electrical and mechanical performance of the DPC-EX was observed in the range of these cycles.

Japan Atomic Energy Research Institute (JAERI) has been developing 13-T, 40-kA conductors for the ITER Central Solenoid Scalable Model Coil. The conductors are composed of high performance [(NbTi).sub.3]Sn strands and titanium conduit. In this development work, 15-km length strand fabrication from a 100-kg billet, chrome plating, cabling of 675 or 768 strands, jacketing of titanium conduit, and so on were conducted by two mass production lines. In this paper, several test results are reported such as critical current measurements of full-scale conductors and AC loss measurements of sub-scale conductors.

The Japan Atomic Energy Research Institute (JAERI) has developed the Advanced Disk and the Hollow Monolithic conductors for the FER/ITER Toroidal Field coils. The Advanced Disk conductor is a Cable-in-Conduit conductor which consists of 324 Nb3Sn strands. The Hollow monolithic conductor has hollow cooling channels and 23 Nb3Sn strands. The JA-FENIX sample consists of a pair of straight legs: one leg is the Advanced disk conductor and another is the Hollow Monolithic one. The FENIX facility at the Lawrence Livermore National Laboratory (LLNL) can provide a magnetic field up to 13T on a sample conductor of over 40cm-length. The performance test of the JA-sample was carried out in Autumn 1992. The critical current, the current sharing temperature, and the stability margin of each conductor were measured in this test. These results are presented and discussed.

The model coil program for the ITER EDA includes the manufacture and testing of one Central Solenoid (CS) model coil and one Toroidal Field (TF) model coil, and test facility preparations. The CS model coil has an inner diameter of 1.6 m and produces a full field of 13 T. The TF model coil is race-track shaped with an outer dimension of 3 m X 4 m. It will be tested in conjunction with an LCT coil to simulate mechanical load conditions. Different kinds of conductors can be tested as inserts to be placed in the bore of the CS model coil. Two facilities at JAERI and KfK can provide ITER-relevant conditions for testing the model coils and inserts. The model coil program will validate the ITER magnet design and the manufacturing feasibility.

A model was developed to elucidate hydrogen recombination coefficients for the front and back side of plasma facing materials. Energetic ions generated imperfection-surface sites and normal-surface sites which were located on the front side. The metal-oxide system was evaluated to obtain the effective recombination coefficient of the back side.

Experimental tests have been conducted by the Japan Atomic Energy Research Institute to characterize the performance of its 30-kA Demo Poloidal Coils. The coils, whose windings are formvar-coated niobium titanium cable-in-conduit conductors, were tested extensively to evaluate their stability margin and current distribution. The results were then used to modify their design and construct new 30-kA conductors.

The Demo Poloidal Coil (DPC) program at the Japan Atomic Energy Research Institute aims to advance superconducting technology to develop superconducting coils for fusion experimental devices. So far, it has designed and tested four kA-class Nb3Sn coils with a 1-m bore. These are the DPC-EX, US-DPC, DPC-TJ and the Toroidal Model Pancake. A report on their operating performance and ac loss measurements is presented.

The Toroidal Model Pancake is a test coil that was constructed by the Japan Atomic Energy Research Institute for fusion reactor studies. Built of a 90-m hollow cooling monolithic conductors, it has an 1-m bore and a 1.5-m outer diameter. It was subjected to ac loss measurements which found the effective diameter of its (NbTi)3Sn and niobium titanium conductors to be 4 micrometer and 39 micrometer, respectively.

The characterization of the Toroidal Model Pancake (TMP) is presented. The TMP is a superconducting coil constructed by the Japan Atomic Energy Research Institute for fusion reactor studies. Built of hollow cooling monolithic superconductors, it begun testing in Aug 1992. Its performance is described along with cooldown, pressure drop, stability and mechanical issues.

The results of tests conducted on the Poloidal Coil (DPC) for a fusion experimental reactor were interpreted. The data was obtained from the DPC's forced-flow-cooled field windings composed of (NbTi)3Sn cable-in-conduit conductors with a current capacity of 24 A. The tests were carried out using transport currents of 15-30 kA and magnetic fields of 2.3-4.5 T. It included another test on a small coil to examine the stability margin during 12-T operation.

The Demo Poloidal Coil-TJ is a superconducting coil for a superconducting fusion magnet. Developed by Toshiba and the Japan Atomic Energy Research Institute, the force-cooled coil is designed for high-magnetic-field and high-current-density operation and is built of Nb3Sn cable-in-conduit conductors. It is electromagnetically and mechanically stable, as borne out by quenching tests and measurements of its critical current and current sharing temperature.