S. Turtu

Test Results of Two European ITER TF Conductor Samples in SULTAN

Four conductor lengths were prepared according to the ITER TF conductor design and assembled into two SULTAN samples. The four lengths are not fully identical, with variations of the strand supplier, void fraction and twist pitch. Lower void fractions improve the strand support and increased twist pitches also lower the strand contact pressure but both tend to increase the AC loss and the lower void fraction also increases the pressure drop so that the mass flow rate in the strand bundle area of the cable is reduced. The assembly procedure of the two samples is described including the destructive investigation on a short conductor section to assess a possible perturbation of the cable-to-jacket slippage during the termination preparation. Based on the DC performance and AC loss results from the test in SULTAN, the impact of the void fraction and twist pitch variations is discussed in view of freezing the ITER conductor design and large series manufacture. A comparison with the former generation of conductors, using similar strands but based on the ITER Model Coil layout, is also carried out. The ITER specifications, in terms of current sharing temperature, are fulfilled by both samples, with outstanding results for the conductor with longer twist pitches.

Test Results of Two ITER TF Conductor Short Samples Using High Current Density Nb

Two short length samples have been prepared and tested in SULTAN to benchmark the performance of high current density, advanced Nb3Sn strands in the large cable-in-conduit conductors (CICC) for ITER. The cable pattern and jacket layout were identical to the toroidal field model coil conductor (TFMC), tested in 1999. The four conductor sections used strands from OST, EAS, OKSC and OCSI respectively. The Cu:non-Cu ratio was 1 for three of the new strands, compared to 1.5 in the TFMC strand. The conductors with OST and OKSC strands had one Cu wire for two Nb3Sn strands, as in TFMC. In the EAS and OCSI conductors, all the 1080 strands in the cable were Nb3Sn. A dc test under relevant load conditions and a thermal-hydraulic campaign was carried out in SULTAN. The CICC performance was strongly degraded compared to the strand for all the four conductors. The current sharing temperature at the ITER TF operating conditions (jop = 286 A/mm2, B = 11.15 T) was lower than requested by ITER.

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The upgrade of JT-60U magnet system to superconducting coils (JT-60SA) has been decided by both parties of Japanese government (JA) and European commission (EU) in the framework of the Broader Approach (BA) agreement. The magnet system for JT-60SA consists of 18 toroidal field (TF) coils, a Central Solenoid (CS) with four modules, seven Equilibrium Field (EF) coils. The TF case encloses the winding pack and is the main structural component of the magnet system. The CS consists of independent winding pack modules, which is hung from the top of the TF coils through its pre-load structure. The seven EF coils are attached to the TF coil cases through supports which include flexible plates allowing radial displacements. The CS modules operate at high field and use Nb3 Sn type superconductor. The TF coils and EF coils use NbTi superconductor. The magnet system has a large heat load from nuclear heating from DD fusion and large AC loss. This paper describes the technical requirements, the operational interface and the outline of conceptual design of the superconducting magnet system for JT-60SA.

Sn Strands

The DEMO plant will demonstrate by mid-century the feasibility of electric power generation by nuclear fusion. In the scope of design studies coordinated by the European Commission, CRPP and ENEA are developing two force flow conductor layouts as candidates for the toroidal field (TF) coils of DEMO. The operating conditions include 82 kA operating current and 13.5 T peak field, placing the DEMO TF conductor at substantially higher performance compared to the ITER TF (68 kA/11.5 T). The innovative winding layout is graded, layer wound with Nb3Sn/NbTi hybridization for both conductor designs, aiming at minimizing the size and the cost of the superconductor: one of the conductors is a wind and react cable-in-conduit conductor with reduced void fraction and has a rectangular shape. The other conductor is a react and wind flat cable with copper segregation and thick conduit assembled by longitudinal weld. The conductor designs were first drafted in 2012 and reviewed in 2013 based on a first round of assessments and on an updated requirements catalog. The manufacture of full-size prototypes is planned in 2014.

Conceptual Design of Superconducting Magnet System for JT-60SA

In the framework of the JT-60SA design activities, EU home team has defined a reference layout for the Toroidal Field conductor: it is a slightly rectangular Cable-In-Conduit NbTi conductor, operating at 25.7 kA with a peak field of 5.65 T. ENEA has assigned LUVATA Fornaci di Barga the task to produce the strands and to perform cabling, whereas jacketing and compaction have been carried out in its own labs. The sample, successfully tested at the CRPP SULTAN facility, has been assembled in such a way as to avoid the bottom joint between the two legs, thus using a single conductor length (about 7 m). An ad-hoc developed solution to restrain the U-bent conductor section (where jacket is not present), consisting in a stainless steel He-leak tight box with an inner structure designed in order to completely block the cable, has been also developed and manufactured by ENEA, where the sample has been also assembled. Instrumentation installation and final assembly of the sample have been performed by the SULTAN team. The main aspects of the sample manufacturing and characterization are here presented and discussed.

Design of Large Size, Force Flow Superconductors for DEMO TF Coils

The conductor for central solenoid (CS) and equilibrium field (EF) coils of JT-60 Super Advanced (JT-60SA) were designed. The conductor for CS is Nb3Sn Cable-In-Conduit (CIC) conductor with JK2LB jacket. EF coil conductors are NbTi CIC conductor with SS316LN jacket. The field change rate (3.9 T/s), faster than ITER generates the large AC loss in conductor. The analyses of current sharing temperature (Tcs)margins for these coils were performed by the one-dimensional fluid analysis code with transient heat loads. The margins of these coils are 1 K for the plasma standard and disruption scenarios. The minimum Tcs margin of CS conductor is 1.2 K at plasma break down (BD). The margin is increased by decreasing the rate of initial magnetization. It is found that the disruption mainly impacts the outer low field EF coil. The disruption decreases the Tcs margin of the coil by >1 K. A coupling time constant of <100 ms, Ni plating, and a central spiral are required for NbTi conductor.

The JT-60SA Toroidal Field Conductor Reference Sample: Manufacturing and Test Results

The DEMO reactor is expected to be the first application of fusion for electricity generation in the near future. To this aim, conceptual design activities are progressing in Europe (EU) under the lead of the EUROfusion Consortium in order to drive on the development of the major tokamak systems. In 2014, the activities carried out by the magnet system project team were focused on the toroidal field (TF) magnet system design and demonstrated major achievements in terms of concept proposals and of consolidated evaluations against design criteria. Several magnet system R&D activities were conducted in parallel, together with broad investigations on high temperature superconductor (HTS) technologies. In this paper, we present the outcomes of the work conducted in two areas in the 2014 magnet work program: 1) the EU inductive reactor (called DEMO1) 2014 configuration (power plant operating under inductive regime) was the basis of conceptual design activities, including further optimizations; and 2) the HTS R&D activities building upon the consolidated knowledge acquired over the past years.

Conductor Design of CS and EF Coils for JT-60SA

The broader approach agreement between Europe and Japan includes the construction of a fully superconducting tokamak, the JT-60 Super Advanced (JT-60SA), as a satellite experiment to ITER. In particular, the whole Toroidal Field magnet system, described in this paper, will be provided to Japan by the EU. All the TF coil main constituents, i.e. conductor, winding pack, joints, casing, current leads, are here presented and discussed as well as the design criteria adopted to fulfil the machine requirements. The results of the analyses performed by the EU and JA to define and assess the TF magnet system conceptual design are reported and commented. Future work plan is also discussed.

Overview of Progress on the EU DEMO Reactor Magnet System Design

The effect of transverse loads on strands has been pointed as a possible cause of the difference observed when scaling transport properties of single strands to those of cable-in-conduit conductors. Single multifilamentary strands inside cables are in fact subject to bending strain due to the electromagnetic forces at operating conditions and to the geometrical layout. Here the influence of pure bending strain, applied in combination with a longitudinal strain, on the critical current of advanced strands for ITER has been studied. The tested samples are single strands inserted inside a thin stainless steel jacket and wound on stainless steel barrels. After the heat treatment, a pure bending strain has been applied transferring the wires on different diameter mandrels, using ad-hoc developed and qualified techniques. Transport critical current has been measured on the single strands before and after the steel jacketing, as well as after the additional application of two different values of maximum bending strain: 0.5% and 0.25%. This was the best choice in order to verify experimentally whether the so-called long twist pitch condition can be applied for the selected strands. The distribution of the bending strain over the strand cross-section has been calculated with finite element numerical codes, and the expected critical current degradation in the limiting cases of short and long twist pitch has been computed and compared with experimental data.

JT-60SA Toroidal Field Magnet System

The Bus Bar III (BBIII), fabricated within the Toroidal Field Model Coil Task of the International Thermonuclear Experimental Reactor (ITER), was tested at the Forschungszentrum Karlsruhe, Germany, in the spring of 2004. The BBIII consists of an approximately 7 m long NbTi dual-channel conductor with a thick square stainless steel jacket, cooled by forced flow supercritical He. It was energized with currents up to 80 kA and operates in its self magnetic field (up to /spl sim/0.8 T). The BBIII was instrumented with Hall-probe heads and arrays, voltage rings and longitudinal voltage taps for electro-magnetic measurements, in order to get experimental data to be used for the validation of a recently developed hybrid thermal-hydraulic electro-magnetic code (THELMA), as well as for the assessment of the possibility of performing a reliable reconstruction of the current distribution in the conductor cross section under controlled conditions. In the tests, current ramps at different rates were applied to characterize the conductor time constants, while two different resistive heaters (one upstream of the BBIII inlet, another one directly on the BBIII jacket) were separately operated in order to approach current sharing in the conductor and to observe the related current re-distribution. In this paper, a summary of the collected experimental results is presented, with particular emphasis on those aspects more relevant for the forthcoming THELMA analysis.