• 한국초전도ㆍ저온공학회논문지
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• 제19권4호
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• pp.31-39
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• 2017

European R&D on designing their version of a DEMO fusion tokamak has recently resulted in the testing of a prototype $Nb_3Sn$ Cable-in-Conduit Conductor (CICC) for the DEMO TF coil. The characteristics and reported results of low temperature performance tests with the prototype CICC sample are compared with those from CICC samples incorporating other recent $Nb_3Sn$ cable designs. The EU-DEMO TF CICC prototype shows performance characteristics similar to that of the ITER CS CICC with short twist pitch. This is a first for a CICC sample that does not have a circular cross section. Assessment of its internal magnetostatic self-field suggests that a reduction in the internal self-field due to the rectangular geometry of the EU-DEMO TF CICC prototype compared to one with a circular geometry may have contributed to the performance characteristics showing current sharing temperature ($T_{cs}$) initially increase then stabilize with repeated electromagnetic loading, similarly to ITER CS CICC results. However, constraints on the internal self-field are not a sufficient condition for this $T_{cs}$ characteristic to occur.

• 대한전기학회:학술대회논문집
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• 대한전기학회 1996년도 하계학술대회 논문집 A
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• pp.37-40
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• 1996

Cable-In-Conduit Conductor(CICC) is widely accepted as an advanced superconductor configuration for large scale applications such as tokamak fusion reactors, MAGLEV (MAGnetic LEVitation), and SMES (Superconducting Magnetic Energy Storage). The stability of CICC cooled with supercritical helium can be very high if it is operated below a certain limiting current. This limiting current can be determined by Stekly type heat balance equation. The stability characteristic of CICC for AC operation is more complicated than that of DC because there are additional instability sources which are associated with local flux change. Ramp-rate limitation is a phenomenon discovered during US-DPC (United States-Demonstration Poloidal Coil) program, which showed apparent quench current degradation associated with high dB/dt. This paper describes recent experimental investigation results on the ramp-rate limitation and discusses current imbalance, induced current, current redistribution due to local quench of the strand in the cable.

• 한국초전도ㆍ저온공학회논문지
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• 제8권2호
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• pp.25-28
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• 2006

The magnet system of KSTAR(korean Superconducting Tokamak Advanced Research) is consisted of 16 TF (Toroidal Field) coils and 14 PF (Poroidal Field) coils. Internal cooling CICC(Cable in Conduit Conductor) type conductor is used for both of TF and PF coil systems. The conduit material for $Nb_3Sn$ cable is Incoloy 908 and 316LN stainless-steel was used as conduit material for NbTi cable. $Nb_3Sn$ CICC is used for all TF coils and PF1-5 coils while NbTi CICC is used for PF6 and 7 coils. $Nb_3Sn$ and NbTi strands were made for KSTAR superconducting strand. They are satisfied with KSTAR superconducotr requirements. The $Nb_3Sn$ strands supplied from three companies; MELCO (Mitsubishi Electric Co.), OAS (Outokumpu Advanced Superconductor) and KAT (Kiswire Advanced Technology) were used. A special CICC jacketing system is developed for the KSTAR CICC fabrication which uses the tube-mill process consisted of forming, welding, sizing and squaring procedures. The. procedures for cabling and jacketing of CICC for TF and PF coils and their results including the geometrical specification and characteristics of strands are described.

• 대한기계학회:학술대회논문집
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• 대한기계학회 2001년도 춘계학술대회논문집A
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• pp.740-745
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• 2001

The plastic deformation behavior of formed CICC fur the superconducting Tokamac fusion device was examined and appropriate manufacturing information was provided. A relation between travel of the bending roller and spring back displacement was obtained via virtual manufacturing. The radius of CICC after forming was expressed as a function of the bend-roll travel. The maximum von Mises stress after spring back was also monitored fur the SAGBO prediction. Next, the variation of the CICC cross-sectional area was examined during the first turn and during conduit bending with the largest curvature. Finally, the coil radius was measured and compared with the data generated from the virtual manufacturing. The measured data showed similar pattern as predicted one. Using the mapping function found to match with the real data, the data from the virtual manufacturing may facilitate accurate manufacturing.

• 대한전기학회:학술대회논문집
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• 대한전기학회 2000년도 하계학술대회 논문집 B
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• pp.801-803
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• 2000

The large scale magnets like thermalnuclear fusion devices are necessary for superconducting CICC cable, When the Cable In Conduit Conductors(CICC) is occurred by the external turbulence, the CICC occurs to quench, The CICC can be broken because the CICC spends all energy in the quench-happened spot. Therefore, it is necessary to develop measurement systems of the quench detection. The measurement systems of the relative good degree of efficiency are the voltage tap sensors. The weak points of voltage tap sensors are effected by EMF noise and inductance. The thermalnuclear fusion devices easily can't measure inductance value because of plasma current. In the experiment, The value of inductance was estimated by FEM techniques and the decrement of Inductance value measured as long as remaining plasma current.

• 한국초전도저온공학회:학술대회논문집
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• 한국초전도저온공학회 2002년도 학술대회 논문집
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• pp.306-309
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• 2002

A superconducting CICC (Cable-In-Conduit-Conductor) is adopted the KSTAR (Korea Superconducting Tokamak Advanced Research) superconducting magnet system which consists of 16 TF coils and 14 PF coils. For the test of KSTAR CICC, an ambient magnetic field of $\pm$ 8 T With a maximum change rate of 20 T/s is required and a background-field magnet system is being developed for SSTF (Samsung Superconductor Test Facility). The CICC for PF1~5 is used as the conductor for background-field coils to check the validity of the PF CICC design. Two pieces of cables have been fabricated and the cable has the length of 870 m and the diameter of 20.3 mm. A continuous CICC jacketing system is developed for the KSTAR CICC fabrication and the jacketing system uses the tube-mill process, which consists of forming, welding, sizing and squaring procedures. The design specification of CICCs and the fabrication process is described.

• 한국초전도ㆍ저온공학회논문지
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• 제5권1호
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• pp.71-75
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• 2003

The conduit fer superconducting cable is welded and plastically deformed during the jacketing process to make the CICC (Cable-in-Conduit-Conductors) fer a fusion magnet. The jacketing process of KSTAR (Korea Superconducting Tokamak Advanced Research) conductors is composed of several sequential steps such as rounding, welding, sizing, and square-rolling. Since the welded zone in Incoloy 908 conduit is brittle and easy to have flaws, there may be a possibility of stress corrosion cracking during the heat treatment of coil when both the induced tensile residual stress and the concentration of oxygen in the furnace are sufficiently high. The steps of the jacketing process were simulated using the finite element method of the commercial ABAQUS code, and the stress distribution in the conduit in each step was calculated, respectively. Furthermore, the variations of residual welding stresses through the steps of the jacketing process were calculated and analyzed to anticipate the possibility of the stress corrosion cracking in the conduit. The concentrated high tensile residual welding stresses along the welding bead decrease by the plastic deformation of the following sizing step. The distribution in residual stresses in the conductor for magnet coil is mainly governed by the last step of square-rolling.

• 한국초전도저온공학회:학술대회논문집
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• 한국초전도저온공학회 2000년도 KIASC Conference 2000 / 2000년도 학술대회 논문집
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• pp.12-16
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• 2000

The mission of SSTF is test of superconducting cables for KSTAR magnets. To make realistic environment for superconductor in SSTF, background magnets are required. Cable-in-conduit conductors (CICC) are widely used for large scale superconducting magnets such as ITER and KSTAR. Main design criteria for conductor of superconducting magnets are stability, operating margin and cable cooling requirement, caused by peak field and the gradient of fields with respect to time, in system. ZERODEE which used energy balance method, is applied for the calculation of stability. To increase conductor performance, three different strands, such as HP-I, HP-II, and HP-III, are tested. The present configuration of CICC is used for main coils of background magnet in SSTF and Central Solenoid coils of KSTAR magnets.

• 한국초전도저온공학회:학술대회논문집
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• 한국초전도저온공학회 2003년도 추계학술대회 논문집
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• pp.298-300
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• 2003

The KSTAR superconducting magnet system consists of 16 TF (Toroidal Field) and 14 PF (Poloidal Field) coils. Both of the TF and PF coil system use internally-cooled Cable-In-Conduit Conductors (CICC). The major achievement in KSTAR magnet system development includes the development of CICC, the development of a full size TF model coil, the development of a background magnetic field generation coil system, the construction of a large scale superconducting magnet. TF and PF coils are in the stage of the fabrication for the KSTAR completion in the year 2005.

• 한국초전도ㆍ저온공학회논문지
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• 제11권3호
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• pp.1-5
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• 2009

It is important to predict AC loss in $Nb_3Sn$ and NbTi cable-in-conduit-conductor (CICC) reliably for the design and operation of large superconducting coils. The hysteresis loss in the superconducting filaments and coupling loss within strands and among strands in a cable or composite are dominant ac losses in superconducting magnets. The coupling loss in a superconductor can be characterized by identifying the coupling constant time $n{\tau}$. To reduce the coupling loss, all the strands (superconductor and Cu) in KSTAR (Korea Superconducting Tokamak Advance Research) are chromium plated with thickness of $l{\pm}0.5{\mu}m$. The ac losses of PF1, PF5 and PF6 coils has been measured by calorimetric method while applying trapezoidal current pulses with various ramp rate from 0.5 kA/s to 2 kA/s. The coupling time constants for $Nb_3Sn$ coils are $25{\sim}55$ ms and the values are not co-related with the coil size, the time constants for NbTi coil is 30 ms.