The Polo coil, a prototype tokamak poloidal field coil, design features and test results

Darweschsad, M.; Heller, R.; Friesinger, G.; Irmisch, M.; Kathol, H.; Komarek, P.; Maurer, W.; Nöther, G.; Schmidt, C.; Schleinkofer, G.; Sihler, C.; Susser, Mervyn; Ulbricht, A.; Wüchner, F.

doi:10.1109/20.508615

Cited by 13 publications

(3 citation statements)

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“…The purpose of the impulse tests is the generation of a voltage distribution similar to the electrical voltage stress under high voltage discharge with a reduced amount of energy. The energy of the impulse was delivered by a capacitor bank of the POLO switching circuit [174]. The impulse was started by firing of ignitrons.…”

Section: Impulse Testsmentioning

confidence: 99%

The ITER toroidal field model coil project

Ulbricht

Duchateau

Fietz

et al. 2005

Fusion Engineering and Design

119

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Section: Impulse Testsmentioning

confidence: 99%

The ITER toroidal field model coil project

Ulbricht

Duchateau

Fietz

et al. 2005

Fusion Engineering and Design

119

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“…For example, high (700 W) heat generation in the Westinghouse coil in LCT [5] (similar to that on T-15, but explained in another way), a ramp rate limit in DPC [6][7][8] and high AC losses with an extremely long decay time for coupling currents in the Experiment on a Single Inner Vertical Coil (EXIV) [9,10] as well as in the 100 k SMES model coil [11]. Only one of the POLO coil test results was successful enough [12].…”

Section: Conductor Design Optimizationmentioning

confidence: 99%

Overview of superconducting techniques for magnetic fusion devices

Ivanov¹

2000

Nucl. Fusion

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The basics of using the cable in conduit conductor (CICC) concept in new projects for magnetic fusion devices as well as the need to improve the CICC design are discussed. Preference is given to a regular inner structure with respect to the usually considered loose one. Optimal solutions for the cooling scheme, the way to avoid all coil pre-assembly tests, the support structure, vacuum vessel (VV) and cryostat design that could provide the best access to the device, load transfer from VV to cryostat and other design features of new projects are discussed.

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“…The Polo coil, an NbTi circular winding with diameter 3 m, was tested in 1994 at the Forschungszentrum Karlsruhe (Germany) [17]. The cable-in-conduit conductor, see figure 1, has two separate hydraulic circuits: stagnant, supercritical He at 4 bar in the annular cable region and forced flow, two-phase He at 4.5 K in the central pipe (mass flow rate 2 g s −1 ).…”

Section: Selection Of Demonstration Coilsmentioning

confidence: 99%

Superconductors for fusion magnets: realizations, prototypes and running projects

Bruzzone¹

1997

Supercond. Sci. Technol.

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This paper gives a review of the application of superconductivity to nuclear fusion magnets. For each machine, a short description emphasizes the special features of the winding technology and conductor design. A selection of demonstration magnets is presented, highlighting the relevant achievements. The evolution of the conductor design in recent decades and the open issues for future research and development activities are discussed.

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The Polo coil, a prototype tokamak poloidal field coil, design features and test results

Cited by 13 publications

References 0 publications

The ITER toroidal field model coil project

The ITER toroidal field model coil project

Overview of superconducting techniques for magnetic fusion devices

Superconductors for fusion magnets: realizations, prototypes and running projects

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