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The optimization of the heat treatment of \Nb3\Sn conductors for toroidal field coils in ITER was attempted to improve the current sharing temperatures (Tcs). Using the strand, we chose the pattern at 570  ^\circ\C for 250 h and 650  ^\circ\C for 100 h as the best, which increased the critical current and maintained the residual resistivity ratio higher than 100. The behavior of the critical current of the strand vs. the magnetic field, temperature, and strain was also improved. This pattern was used on two conductors, and their performances were tested. Tcs was evaluated over 1000 electromagnetic cycles and one thermal cycle. A sharp Tcs degradation occurred at 50 cycles. Then Tcs decreased linearly. Although this tendency was similar to the conductors that were heat treated with the original pattern, the degradation rates were improved. The ac losses (Q) before cycling were approximately 10% lower than those of the original pattern. Q after cycling became almost equivalent between two patterns. The conductor was inspected after the test, which showed that the conductor under the high-magnetic-field zone had contracted by approximately 600 ppm during the test. Some clearly deformed strands were observed under the high-magnetic-field zone, which could degrade Tcs. [ABSTRACT FROM PUBLISHER]

The performance of four \Nb3\Sn conductors for the ITER central solenoids was tested. The current sharing temperatures (Tcs) were measured over approximately 9000 electromagnetic cycles, including two or three thermal cycles between 4.2 K and room temperature. Tcs increased and became almost constant through the cycling. The gradient of the electric field against the temperature gradually decreased against cycling. The degradations caused by the electromagnetic force of the short twist pitch conductors were smaller than that of the original twist pitch conductor. The ac losses of short twist pitch conductors were several times higher than that of original twist pitch conductor. The dents and the removals of the Cr plating on the strands, which were formed during cabling, decreased the electric resistance between strands, which may cause the observed high ac loss. Inspection of the cable showed neither a clear bias of cable in the cross-sectional surface nor distorted strands in the lateral face. The high rigidity of the short twist pitch cable could prevent these plastic deformations, caused by the Lorentz force. [ABSTRACT FROM PUBLISHER]

The performances of six Nb3Sn conductors for the ITER Toroidal Field coils were tested. Four of them showed similar degradation rates of their current sharing temperatures Tcs over 1,000 electromagnetic cycles. By contrast, two of them showed sharp Tcs degradations at 50 cycles, after which their slopes became similar to those of the other four conductors. These two cables seemed to shrink under high magnetic fields during the first 50 cycles, which caused the sharp Tcs degradation. This shrinkage might arise from a decline in cable rigidity due to, for example, the deformation of strands or the breakage of the Nb3Sn filaments. The four mass-produced conductors had roughly the same AC loss before cycling. After 1,000 cycles, the AC losses of all the conductors decreased markedly to less than half of those before cycling, and the values became approximately the same. After the test campaign, the destructive inspection of two of the conductors made it clear that the conductor had shrunk by about 520 ppm under the high magnetic field during the test. It was also clarified that some strands were visibly deformed under the high magnetic field, whereas those under the low magnetic field did not look distorted. This plastic deformation of the strands could be one of the major reasons for the Tcs degradation with cyclic operation. [ABSTRACT FROM PUBLISHER]

Japan Atomic Energy Agency (JAEA) was the first to start the mass production of the TF conductors (jacketing) in March 2010 among the 6 parties who are procuring TF conductors in the ITER project. [ABSTRACT FROM PUBLISHER]

The magnet system for ITER comprises 18 Toroidal Field (TF) Coils using Nb3Sn cable-in-conduit superconductor, which operate at 4.5 K in supercritical helium. The procurement of the TF Coils and Structures is amongst the first which have been launched following the creation of the ITER Organization (IO). It is organized in 4 phases. A Procurement Design Readiness Review held in April 2008 confirmed the readiness of the design to proceed with Phases I and II. Procurement Arrangements (PA) were signed with the European and Japanese Domestic Agencies (DA) respectively in June and November 2008. After a brief description of the TF Coils and Structures, the paper gives an overview of the PA showing the milestones towards series production. The procurement strategy of both DA involved is described, in particular the first step which covers pre-production activities: qualification of raw materials, manufacturing trials, mock-ups and full-scale prototype radial plates, impregnation tests and, possibly, winding trials. The work carried out by IO is also presented: optimization of the cover plate welding to satisfy the allowable stress criteria while minimizing the associated distortions, qualification of blends of cyanate ester with epoxy resin for the impregnation of the winding packs and design of the coil terminal region including integration of the needed instrumentation. [ABSTRACT FROM AUTHOR]

The authors performed trials and studies to solve the remaining technical problems for manufacturing the ITER Toroidal Field (TF) coil. Cover plate (CP) welding deformation of a side double pancake may achieve specified tolerances with correction of a long wave distortion through the use of a small load, although complicated welding operations are required. In addition, JAEA developed a new procedure for manufacturing a CP at a reduced cost. Straight and curved CPs are fabricated by hot-rolling and cold-drawing and bending the straight CP with sufficient accuracy. Applying these results, JAEA has started procurement of the TF coil, beginning with qualification trials. Manufacture of a full-scale regular RP, finalization of a manufacturing plan and drawings and design of tooling will be completed by the middle of 2011. [ABSTRACT FROM AUTHOR]

Coupling current losses in large scale Cable-In-Conduit Conductors (CICCs) for fusion apparatuses are sometimes annoyance because nobody can estimate how large the coupling loss is before fabricating hundred-meters of long conductor. Our approach for analyzing and estimating the loss based on real strand traces is unique compared with that of other research groups. In process of this approach, we have already measured the traces in two types of CICCs, one has circular cross section and the other has rectangular one. The cabling pattern of the former is 34 = 81 the latter is 34 x 6 = 486. The flux linkage area in a coupling current loop which consists of two contacting strands is evaluated as an indicator of driving force of the coupling current, which is proportional to the square root of AC loss per unit time. The flux linkage areas showed that the losses of rectangular-shape CICC would be larger than that of circular-shape CICC. [ABSTRACT FROM AUTHOR]

In this paper we report the main test results obtained on the Poloidal Field Conductor Insert coil (PFI) for the International Thermonuclear Experimental Reactor (ITER), built jointly by the EU and RF ITER parties, recently installed and tested in the CS Model Coil facility, at JAEA-Naka. During the test we (a) verified the DC and AC operating margin of the NbTi Cable-in-Conduit Conductor in conditions representative of the operation of the ITER PF coils, (b) measured the intermediate conductor joint resistance, margin and loss, and (c) measured the AC loss of the conductor and its changes once subjected to a significant number of Lorentz force cycles. We compare the results obtained to expectations from strand and cable characterization, which were studied extensively earlier. We finally discuss the implications for the ITER PF system. [ABSTRACT FROM AUTHOR]

Coupling loss with long time constants has been found a troublesome phenomenon for large size magnet application of superconductor because it would not be simply estimated from AC loss measurement of short sample conductor. In order to investigate the mechanism of the loss, we measured trajectories of strands of sample CIC (34 = 81) conductor. The measured length is 1 m along the conductor axis. By analyzing those trajectories, two important facts are cleared. One is that contact periods between two strands are calculated by the function of twisting pitches, not always by the Least Common Multiplier of twisting pitches. The other is that contact probabilities of two strands at each contactable point of sub cables depend on the difference of rotation angles of sub cables. To confirm the validity of this method, we calculated lengths of coupling current loops within 1 m in length by using experimentally obtained contact probabilities. The results are in very good agreement with lengths obtained from experimental results of strand trajectories. Then we computed the loop length in the long conductor (<100 m), it was obtained that the average loop length would reached about 3 m. [ABSTRACT FROM AUTHOR]

New technical issues are initiated in the ITER Toroidal Field (TF) coil manufacturing because of the scale-up by about 3 times from the TF model coil. The major issues are feasibilities of high accuracy, automatic winding, precise prediction of conductor elongation after heat treatment to insert the conductor in narrow groove of a radial plate, and development of insulation system, which can sustain irradiation of a neutron fluence of 1022 n/m2 and, in addition, be impregnated in a suitable duration. One of major parts of automatic winding machine, a bending roller head, was developed and trial winding using short sample conductor was successfully carried out. The elongations of the TF conductor after the heat treatment is evaluated to be in ±0.03% In addition, results of impregnation test using short samples indicate that it will take about 60 h to impregnate TF coil double-pancake by the method developed in the model coil project. New impregnation technique is under development to shorten the impregnation period. [ABSTRACT FROM AUTHOR]

A new generation of ITER TF conductor samples has been assembled and tested in SULTAN in 2007 following a common procedure agreed among the ITER parties. The test results of six SULTAN samples, made of twelve conductor sections manufactured in Europe, Japan, Korea and Russia, are reported here. The conductor layout reflects the ITER TF conductor design, with minor differences for the Nb3 Sn strand characteristics, void fraction and twist pitch. The object of the test is a straight comparison with the ITER requirement of 5.7 K current sharing temperature at 68 kA current and 11.3 T field. A broad range of behavior is observed. [ABSTRACT FROM AUTHOR]

AC losses with long time constants can not be simply estimated from a short sample conductor because there are many irregular loops formed by strands strongly displaced from their original positions. In our previous work, we measured trajectories of 81 strands of NbTi conductor and it was proved that strongly displaced strands produced many line contacts with other strands, and thereby caused low contact resistance and long time constants. Long loops due to the displacement of strands should also produce large AC loss because the time constant of the loss is proportional to the inductance, i.e., the length of coupling current loops. In order to investigate the long loops in practical conductors, we developed a method to estimate the strand positions over the entire length. In this method, only one cross section of the conductor is required to calculate gravities of each sub-cable. The strand trajectories are obtained in a manner that the same order sub-cables rotate around the gravity to form one order higher sub-cable. The estimated trajectories are in good agreement with the measured ones, with errors of 1 mm. [ABSTRACT FROM AUTHOR]

In this paper we discuss the predictive analysis performed in support of the test program of the International Thermonuclear Experimental Reactor (ITER) Poloidal Field Conductor Insert (PFCI). A subset of the test program items was considered, with particular emphasis on DC performance and AC losses. The results and implications of the comparison of selected predictions from different laboratories will be presented. [ABSTRACT FROM AUTHOR]

Effects of tensile and compressive strain on critical currents (Ic) of Nb3Al strand and cable-in-conduit (CIC) conductor were investigated using a newly developed apparatus. The Nb3Al strand manufactured by jelly roll process was 0.74 mm in diameter with a copper/noncopper ratio of 4.05. The CIC conductor sample which consists of two Nb3Al strands and one copper wire inserted into a stainless steel conduit was prepared. The effective strain and magnetic field ranges for the CIC conductor sample were from 0.91% to +0.26% and from 6 T to 11 T, respectively, it is clearly shown that the Ic decrease of the CIC conductor sample by longitudinal strain is relieved considerably compared with the strand sample, indicating the relaxation of intrinsic strain in Nb3Al filaments in CIC conductor due to the cabling effect. The relaxation effect is about +0.1% for -0.8% in effective strain of CIC conductor. [ABSTRACT FROM AUTHOR]

Cryogen-free superconducting magnets are becoming popular due to their simple operation compared with the conventional liquid Helium cooled magnet. However, for higher fields such as those greater than 14 T, the cryogen-free magnet has not become popular yet, because it is difficult to design and manufacture since the critical current is markedly reduced at higher fields and higher temperatures. We have developed two types of 15 T cryogen-free superconducting magnets, which will be among the highest field magnets in operation as cryogen-free with a GM cryocooler. One magnet has a 52 mm room temperature bore with overall dimensions of 820 mm in diameter and 680 mm in height. This magnet is designed so as to be simple in both operation and installation. Therefore, the magnet is cooled with a single 1 W GM cryocooler. Also, all coils are connected in series and charged with a single power supply. The magnet was successfully charged up to 15 T in 30 minutes and then charged slowly to 16 T without quench. The other magnet has a 170 mm room temperature bore with overall dimensions of 980 mm in diameter and 1015 mm in height. This magnet is designed so as to be used with temperature controlled sample cryostats with an outer diameter of 168 mm. The magnet is cooled with four 1 W GM cryocoolers and charged with a single power supply. The magnet was successfully charged up to 15 T in 90 minutes. Both magnets will enable high magnetic field research or application with very easy operation. [ABSTRACT FROM AUTHOR]

As the test of the PFCI is foreseen in 2006 at JAERI Naka, Japan, it is essential to consider in detail the lessons learned from the short NbTi sample tests, as well as the issues left open after them, in order to develop a suitable test program of the PFCI aimed at bridging the extrapolation gap between measured strand and future PF coil performance. Here we consider in particular the following issues: 1) the actual possibility to quench the PFCI conductor in the Tcs tests before quenching the intermediate joint, 2) the question of the so-called sudden or premature quench, based on SULTAN sample results, applying a recently developed multi-solid and multi-channel extension of the Mithrandir code to a short sample analysis; 3) the feasibility of the AC losses calorimetry in the PFCI. [ABSTRACT FROM AUTHOR]

AC losses consist of both regular losses that are proportional to cable twisting pitch squared and irregular losses that could not be estimated from short conductor sample test results. It was explained from our previous works that irregular loops in conductor which are caused by asymmetric strand positions as a result of low void fraction of CIC conductor, produce the losses with long time constants up to several hundred seconds. The observed long time constant indicates that the typical loop length should be about LCM (Least Common Multiplier) of all sub-staged cable pitches, and that contact conditions between the two strands forming the loop should be line contact. In order to investigate the contact conditions in detail, we traced 81 (= 3 × 3 × 3 × 3) strands every 11 mm of CIC sample conductor with I m in length whose strands are NbTi/Cu without any surface coating. The measured traces of 81 strands show that asymmetric strand positions, in other words, large displacements of strands from their original positions due to compressing the conductor provide many line contacts. It is found that the averaged line contact length reaches about 10 mm that is three order of magnitude larger than the 10-2 mm of point contact length. [ABSTRACT FROM AUTHOR]

Japan Atomic Energy Research Institute (JAERI) has studied manufacturing processes of radial plates used in Toroidal Field coils of the International Thermonuclear Experimental Reactor (ITER). A manufacturing procedure, in which grooved radial plate blocks are connected by laser welding, was considered as the most rational one. A trial fabrication of a radial plate which has the same dimensions of the actual ITER was performed to clarify the feasibility of this procedure through the measurements of deformation during machining and welding. As a result, it was concluded that this manufacturing procedure could be used for the ITER because it could achieve the required tolerances for manufacturing of the radial plates. [ABSTRACT FROM AUTHOR]

A quench test of a 13 T and 46 kA Nb3 Al cable-in-conduit conductor was performed. Thermo-hydraulic quench back (THQB) was observed clearly for the first time in a large magnet when the temperature was set at 0.5 K lower than the current sharing temperature and normal zone propagation velocity reached 70 m/s. These results suggest possibility of the THQB should be considered in a design of a large magnet system in case that the operating temperature is considerably close to the current sharing temperature. Calculation results by one-dimensional stability and quench simulation code are compared with the experiment. Although the calculation results show some discrepancy with the experiment they support the conclusion that the THQB took place. Some improvements of the model seem necessary. [ABSTRACT FROM AUTHOR]

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, ∼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 ∼6 T, and it will be cooled by supercritical He at ∼4.5 K and ∼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, pickup 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. [ABSTRACT FROM AUTHOR]

The world's largest high-field pulsed super conducting magnet, the CS Model Coil, was manufactured and tested by the International and Participant Teams for the verification of new design principles and fabrication concepts proposed for the ITER CS. State-of-the-art technologies were developed through the CS Model Coil program including high-performance Nb3Sn strand, 46 kA cable-in-conduit conductor, and coil fabrication using "wind, react, insulate and transfer" method that enables Nb3Sn to be applicable to a large magnet with mechanically rigid and high-voltage insulated windings. The CS Model Coil reached its design point of 13 T and 46 kA and demonstrated the pulsed operation of 0.6 T/s to 13 T, achieving all the development goals. The experiments of the CS Model Coil and three single layer Inserts have provided detailed knowledge on the performances of the coils and conductors including the hydraulic behavior, AC losses and critical currents. In particular, significant effects of large electromagnetic forces acting on the conductor were observed in these performances and the movement of cable inside the conductor was identified to be important. Results and experiences from the program are not only useful for the design optimization of the ITER magnets, but also applicable to other areas such as high energy physics. [ABSTRACT FROM AUTHOR]

Using a 20 kA Nb3 Al cable-in-conduit (CIC) conductor, a two-turn fl-shaped test coil with a height of 2.0 m and a width of 1.8 m was fabricated by a react-and-wind (R&W) method. After heat treatment, the conductor was formed to the fl-shaped coil. Bending strain of the conductor is ±0.4% at critical current (Ic) test section. The Ic exceeded 31.4 kA at 7.3 T, 4.2 K. In spite of the stainless steel conduit, a critical current density (Jc) of 89% of the strand value was attained. The total strain of the fl-shaped test coil was about -0.57%, the same as the triplex-CIC sample fabricated by a W&R method, indicating that bending did not influence the 4. The 4 at 7.4 T, 5.6 K including the margin of 1 K exceeds the operational current Therefore, it was concluded that the R&W method is quite appropriate for the toroidal field coil of JT-60SC. [ABSTRACT FROM AUTHOR]

Development of Nb3Sn cable-in-conduit (CIC) conductors with stainless steel (SS) jackets for the central solenoid (CS) and divertor coil of the superconducting tokamak called JT-60SC, is presented. Full-size and subsize conductors were manufactured using two types of high-copper-ratio and high-Jc strands made by different processes (bronze and internal-tin). Use of a SS jacket causes decrease of critical current of the conductor because compression strain due to difference of the thermal contraction between the SS jacket and Nb3Sn cable is generated. The strains are estimated from the results of critical current measurements using short full-size samples which were pool-cooled. In order to investigate Jc - B - T characteristics of a CIC conductor, a novel apparatus, which can control the temperature and magnetic field, has been developed. This apparatus makes it possible to measure directly the critical current and temperature of a subsize conductor, which is composed of two superconducting strands, one copper wire and a SS jacket. Strains of the subsize conductors are estimated from the results of the J3 - B - T characteristic measurements. By comparison of the estimated strains between the full-size and subsize samples, we concluded that the characteristics of the full-size conductor can be predicted by measuring of J3 -. B - T characteristics of subsize conductors. Finally, the strain of the actual conductor under the operational condition is estimated. The result confirms that the developed conductors are applicable to CS and divertor coil of JT-60SC. [ABSTRACT FROM AUTHOR]

Cable-in-conduit conductors (CICC) consisting of about 1000 Nb3 Sn strands with an outer diameter of about 0.8 mm, have been designed for the TF and CS coils of the ITER. The rated currents of these coils are 68 and 40 kA, respectively. Two joint types (Butt and Lap) were developed during the CS Model Coil project. The performance of these joints was evaluated during the operating tests and satisfactory results were obtained. The joints of the TF coils are located outside of the winding in a region where the magnetic field is about 2.1 T, a very low value as compared to the maximum field of 11.81 at the winding. The CS joints are located at the coil outer diameter and embedded within the winding pack to minimize the space occupied around the coil. The maximum fields at the CS joint and winding are 3.5 and 13 T, respectively. For the TF coils and the CS, the joints are cooled in series with the conductor at the outlet. The maximum allowable temperature increase due to the joule heating in the joints is 0.15 K to reduce the heat load on the refrigerator. The stability of these joints at the ITER operating condition was analyzed by using the experimental data. The analysis shows that both joint types are applicable to the full size ITER coils. [ABSTRACT FROM AUTHOR]

The ITER Central Solenoid (CS) is 12 m high and 4 m in diameter. The CS consists of a stack of 6 electrically independent modules to allow control of plasma shape. The modules are compressed vertically by a p re-compression structure to maintain contact between modules. The CS conductor is CIC conductor with Nb3 S ii strands and a steel conduit. The CS model coil and insert coil test results have shown that the conductor design must be modified to achieve an operation margin. This required either to increase the cable diameter or to use strand with a higher current capability. A bronze-process (NbTi)3Sn strand is proposed to achieve a higher critical magnetic field. A square conduit with a high Mn stainless steel is proposed as it can satisfy fatigue requirements. The inlets are in the high stress region and any stress intensification there must be minimized. The preload structure is composed of a set of tie-plates, flanges, buffer plates, and wedges. These design proposals satisfy all ITER operational requirements. [ABSTRACT FROM AUTHOR]

Japan Atomic Energy Research Institute has developed a low carbon and boron added 0.03C-22Mn-13Cr-9N1-1Mo..0.24N-0003B steel (JK2LB) for a conductor jacket of the central solenoid for the International Thermonuclear Experimental Re- actor in collaboration with Kobe Steel Ltd. The feature of JK2LB is to have enough ductility and toughness after reaction heat treatment to produce Nb3 Sn. Tensile properties, fracture toughness, and fatigue crack propagation rate were measured at 4 K by using samples taken from an intermediate billet and the final jacket of JK2LB, which were produced with a mass production line. The average elongation and fracture toughness at 4 K after the heat treatment are 33% and 93 Mpa&radicm; for the final jacket, and 30% and 121 Mpa&radicM; for the intermediate billet, respectively. The mechanical properties of weld metals were also evaluated and all data satisfied the ITER targets. It was also clarified that improvement of ductility and toughness were caused by reduction of chromium carbide precipitations due to low carbon and boron addition. [ABSTRACT FROM AUTHOR]

The Central Solenoid Model Coil (CSMC) was designed and built by an ITER collaboration in 1993-2001. Three heavily instrumented Inserts have been also built for testing in the background field of the CSMC. The Nb[sub 3]Al Insert was designed and built by Japan to explore the feasibility of an alternative to Nb[sub 3]Sn superconductor for fusion magnets. The Nb[sub 3]Al Insert coil was tested in the CSMC Test Facility at the Japan Atomic Energy Research Institute, Naka, Japan in March-May 2002. It was the third Insert tested in this facility under this program. The Nb[sub 3]Al Insert coil was charged successfully without training in the background field of the CSMC to the design current of 46 kA at 13 T peak field and later was successfully charged up to 60 kA in 12.5 T field. This paper presents the test results overview. [ABSTRACT FROM AUTHOR]

The Central Solenoid Model Coil (CSMC) was designed and built by ITER collaboration between the European Union, Japan, Russian Federation and the United States in 1993-2001. Three heavily instrumented insert coils have been also built for testing in the background field of the CSMC to cover a wide operational space. The TF Insert was designed and built by the Russian Federation to simulate the conductor performance under the ITER TF coil conditions. The TF Insert Coil was tested in the CSMC Test Facility at the Japan Atomic Energy Research Institute, Naka, Japan in September-October 2001. Some measurements were performed also on the CSMC to study effects of electromagnetic and cooldown cycles. The TF Insert coil was charged successfully, without training, in the background field of the CSMC to the design current of 46 kA at 13 T peak field. The TF Insert met or exceeded all design objectives, however some interesting results require thorough analyses. This paper presents the overview of main results of the testing—magnet critical parameters, joint performance, effect of cycles on performance, quench and some results of the post-test analysis. [ABSTRACT FROM AUTHOR]

In 2000, Japan Atomic Energy Research Institute (JAERI) and its collaboration team accomplished many kinds of experiments under the magnetic field of 13T for the ITER Project. The target coils are the central solenoid (CS) model coil and the CS insert coil. In 2001, the test using both the CS model coil and the toroidal field (TF) insert coil was carried out and successfully finished. During the experiment, we have measured the change of amount of mechanical disturbances inside the coil using acoustic emission (AE) technology. In this paper, we report the general trend of AE characteristics obtained in the experiments for two years. That is to say, as for the CS model coil, we investigated the training characteristics of the CS model coil that experienced one cooling cycle from 4.2 K to room temperature. As a result, we confirmed the training effect of the CS model coil wound by forced flow CIC conductors. On the other hand, as for the insert coil, some peculiar AE signals were observed during the CS insert coil cyclic test. On this matter, a re-examination was carried out. [ABSTRACT FROM AUTHOR]

Evaluation method of critical current (l[sub c]) and current sharing temperature (T[sub cs]) is proposed and applied to the experiment of ITER Center Solenoid (ITER-CS) Model Coil Insert, which is a Nb3Sn superconducting coil. Voltage behavior related to normal state transition of conductors during I[sub c] or T[sub cs] measurement is not yet well understood especially in case of such a large cable with more than one thousand strands as ITER-CS Insert, because magnetic field, which makes large effect on superconducting property, is not constant inside of a cable. From the detailed analysis on the voltage behavior of coils, it is found that integral of electric field averaged over conductor cross section along strand-longitudinal direction is equal to the voltage which is measured by voltage taps during a coil test. It is because twist pitch of a cable is less than the range of longitudinal field variation in case of a large-cable-conductor coil. This evaluation method can estimate voltage behavior and predict I[sub c] and T[sub cs] values, which are important parameters for the design of a large-cable-conductor coil, based on the property of strands composing the conductor. [ABSTRACT FROM AUTHOR]

Examines the performance of the International Thermonuclear Experimental Reactor-Central Solenoid superconducting coil and insert. Ramp up rate in pulsed operation; Quench current of the coils; Critical current and the limiting current.

Reports on tests performed on the International Thermonuclear Experimental Reactor (ITER) Central Solenoid (CS) Model Coil and CS Insert. Building of the inner and outer modules of the CS model coil to demonstrate the feasibility of a superconducting CS for ITER and other tokamak reactors; Critical current tests; Current losses; Cooldown and thermohydraulics.

Describes the instrumentation of the Central Solenoid Model Coil (CSMC) and the CS Insert. Building of the CSMC by the United States and the Japanese home teams as part of an international collaboration with the European Union and Russia; Description of the CS model coil and CS insert coil; Instrumentation map of strain gauge, displacement gauge and acoustic emission sensor.

Describes the status of the structural and construction features of the International Thermonuclear Experimental Reactor (ITER) Model Coil Program and test facilities for the model coils. Central model coil and insert coils; Toroidal model coil; Model coil test facility; Conclusions.

The residual resistivity of copper stabilizer (/spl rho/), is deteriorated by the diffusion of chromium plated on the strand surface during heating of Nb/sub 3/Al strands. To improve this degradation, the copper stabilizer has been located on the center of the cross-section of the strand rather than near the strand surface. As a result, /spl rho/ is improved to one-sixth of previous value achieving 1.6/spl times/10/sup -10/ /spl Omega/m in a laboratory scale strand. The critical current density (J/sub c/) of this strand was 715 A/mm/sup 2/ at 12 T. This technique was applied to prototype strands to be used in a Nb/sub 3/Al Insert Coil developed in ITER R&D programme resulting in a /spl rho/ of 1.5/spl times/10/sup -10/ /spl Omega/m, which meets ITER specification. In addition, the improvement of J/sub c/ was also studied by two-stage reaction including a continuous heating procedure, which resulted in a J/sub c/ of 1130 A/mm/sup 2/ at 12 T and good longitudinal uniformity. [ABSTRACT FROM PUBLISHER]