The Westinghouse coil is one of three forced-flow coils in the six-coil toroidal array of the International Fusion Superconducting Magnet Test Facility at Oak Ridge National Laboratory. It is wound with an 18-kA, NbaSn/Cu. cable-inconduit superconductor structurally supported by aluminum plates and cooied by 4-K, 15-atm supercritical helium. The coil is instrumented to permit measurement of helium temperature, pressure, and flow rate; structure temperature and strain; field: and normal zone voltage. A resistive heater has been installed to simulate nuclear heating, and inductive heaters have been installed to facilitate stability testing. The coil has been tested both individually and in the six-coil array. The tests covered charging to full design current and field, measuring the current-sharing threshold temperature using the resistive heaters, and measuring the stability margin using the pulsed inductive heaters. At least one section of the conductor exhibits a very broad resistive transition (resistive transition index-4). The broad transition, though causing the appearance of voltage at relatively low temperatures, does not compromise the stability margin of the coii, which was greater than 1.1 J/cm 3 of strands. In another, nonresistive location, the stability margin was between 1.7 and 1.9 J/cm 3 of strands. The coil is completely stable in operation at 100% design current in both the single-and six-coil modes.
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