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A number of models for the critical surface of Nb3 Sn, and in general A15 superconductors, have been developed in the past years. This paper compares the most common parameterizations using consistent notation. Although the parameterizations appear dissimilar at first sight, they are in reality all based on a fit of the normalized pinning force vs. the reduced field, and have similar scalings for the critical field and critical temperature based on a Unified Scaling Law. In this paper we take the various parameterizations as a basis for a generic scaling proposed for the characterization and production follow-up of the ITER Nb3 Sn strands. The accuracy of the scaling is estimated using the fitting residuals on various sets of Ic (B, T, ϵ) data available in literature. We discuss the results, and give our view of the work towards a unified, practical parameterization. [ABSTRACT FROM AUTHOR]

The measurement of the current sharing temperature of a Cable in Conduit Conductor is a complex task. The voltage traces measured on the conduit (jacket) of these conductors during tests, give V∙I characteristics that significantly differ from the typical transition from the superconducting to the resistive state measured on single strands. In these measurements, after an initial ramping of the transport current up to the test value, the temperature is increased by steps until the resistive transition occurs. However, even in the first phase of the current ramping, when the cable is still fully or almost superconductive, early voltages are measured along the jacket at different angular positions around the cable. In particular, this was observed in the ITER TF conductors recently tested in the SULTAN facility at Villigen PSI, Switzerland. These samples showed significant ramps of voltages taken in the proximity of the joints and terminations. The present paper gives a possible qualitative and quantitative explanation for this mechanism. Numerical simulations are also shown which reproduce the scattering of the voltages measured around the jacket at a given location, showing the effect of the voltage taps position along the jacket in a six sub-cable model. [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]

For the superconducting magnet system of the International Thermonuclear Experimental Reactor, ITER, 60 current leads for a total current of more than 2500 kA are needed. The use of High Temperature Superconductor current leads (HTS-CL) will reduce the resultant large refrigerator load considerably. Within the EU Fusion Technology Programme, the Forschungszentrum Karlsruhe and CRPP have developed and built a 70 kA HTS-CL demonstrator optimized for 50 K Helium operation. In 2004-2005, the CL was successfully tested with 50 K as well as 80 K Helium, and finally with LN2. In the meantime, ITER has decided to use HTS-CLs for the magnet system and China has taken the responsibility to provide the sc feeders for ITER including the current leads. In this paper, the performance of a 70 kA HTS-CL in fault conditions will be analysed using the FZK-CRPP demonstrator as a model. In particular, the quench performance and the safety behavior in case of a loss-of-flow accident (LOFA) will be described and an extrapolation of the leads performance in real ITER conditions will be given. Finally the effect of different shunt materials is discussed. [ABSTRACT FROM AUTHOR]

The ITER Organization (10) and the Institute of Plasma Physics at the Chinese Academy of Sciences (ASIPP) are jointly developing the 68 kA current leads using High Temperature Superconductors for the superconducting Toroidal Field (TF) magnet system of the International Thermonuclear Experimental Reactor, ITER. The proposed design consists of a conventional helium cooled heat exchanger operating between 65 K and 320 K and an HTS module covering the low temperature end using Bi-2223 tapes. The first HTS current lead prototypes will be tested in the EAST tokamak facility at ASIPP. This paper discusses the design of the first, 68 kA, TF HTS lead, which is the main emphasis of this first stage of the ITER current lead development program. [ABSTRACT FROM AUTHOR]

For the superconducting magnet system of the International Thermonuclear Experimental Reactor, ITER, 60 current leads for a total current of more than 2500 kA are needed. To reduce the resultant large refrigerator load at 4.5 K, High Temperature Superconductor current leads (HTS-CL) could be used. Therefore, EFDA CSU Garching had launched a development program for a 70 kA HTS-CL demonstrator. The Forschungszentrum Karlsruhe and CRPP developed and built this CL optimized for 50 K Helium operation. In 2004, the CL was successfully tested in the TOSKA facility at the Forschungszentrum Karlsruhe. The very encouraging results lead to testing this CL with 80 K Helium because ITER provides a large 80 K Helium cooling capacity for the thermal shields. At the end of last year, the test could be successfully performed demonstrating that high current capacity current leads can be stably operated at about 80–85 K. Recently, the CL was retested using liquid nitrogen which would be an interesting alternative option. In this paper, the test results for the 80 K He cooling operation as well as for the operation with LN2 are presented and compared to the results obtained before for nominal conditions. [ABSTRACT FROM AUTHOR]

Numerous manufacturers and different strand processing techniques are involved with the production of the Nb3 Sn strand material required for ITER. The superconducting transport properties of brittle Nb3 Sn layers strongly depend on their strain state. Hence, the thermal compression and the substantial transverse load in combination with the key choice for the cabling pattern of the CICCs, will determine their performance. Knowledge of the influence of axial strain, periodic bending, and contact stress on the critical current (Ic) of the used Nb3 Sn strands is inevitable to gain sufficient confidence in an economic design and a stable operation of ITER. We have measured the Ic and n-value of Nb3 Sn strands from various manufacturers in the TARSIS facility, when subjected to spatial periodic bending and contact stress. The Ic and n-values have been determined for applied axial compressive and tensile strain varying from -0.8% up to +0.5%, between T = 4.2 K and 10 K and B = 6 T to 14 T. The strain sensitivity varies appreciably for different strand types. We present a selection of the results obtained so far. [ABSTRACT FROM AUTHOR]

Describes the design of the International Thermonuclear Experimental Reactor (ITER) magnet and model coils. Key features of the ITER Final Design Report magnet design; Magnet design drivers for Reduced Technical Objectives/Reduced Cost-ITER; Characteristics of the model coils.

Assesses the radial plate design of the Toroidal Field superconducting coil winding pack of the International Thermonuclear Experimental Reactor. Design options for the winding pack; Features of the radial plate design; Effective elastic characteristics of the winding pack; Stress analysis.

Focuses on the use of the pancake-wound technique in the development of the central solenoid (CS) coil of the Reduced Technical Objective/Reduced Cost for the International Thermonuclear Experimental Reactor. Coil design; Two alternatives for the CS conduit reinforcement; Assessment of the CS structure performance through an axisymmetric finite element method.

Evaluates the performance of the coils made of Cable-In-Conduit Nb3Sn conductors for the Central Solenoid of the International Thermonuclear Experimental Reactor. Performance of fast ramp tests; Distribution of the current inside the cable; Measurement of the critical current; Determination of the fastest possible current ramps.

Focuses on the development of an Nb3Sn conductor joint for the Central Solenoid (CS) Model Coil of the International Thermonuclear Experimental Reactor. Configuration of the butt joint for the CS Model Coil; Fabrication of hairpin cable; Joint resistance; Measurement of the quench current in various temperatures and fields.

Describes the design and fabrication of an Central Solenoid Model Coil developed by the International Thermonuclear Experimental Reactor. Overview of the inner module fabrication; Peak field produced by the model coil; Coil winding procedure; Insulation technology and materials.