The quench protection in high-temperature superconducting (HTS) magnets is a well-known issue. Therefore, it is fundamental to have reliable models available for the analysis of the quench propagation in such magnets. The quench propagation in an ENEA HTS cable-in-conduit conductor, developed for fusion applications with the particular reference to an insert for the central solenoid of the Italian Divertor Tokamak Test (DTT) facility, is analyzed here with a new 1-D multiregional thermal-hydraulic and electric model, based on dedicated preliminary analyses and experimental and numerical characterization of the conductor. It is shown that large temperature differences arise in the conductor cross section during the quench propagation. The model is then applied to parametrically assess the effects of delay and fast current-discharge times on the conductor peak temperature, to avoid damaging the HTS. The parametric study shows that both the delay and current discharge time should stay below 0.5 s, to keep the peak temperature below 150 K.

The ITER Central Solenoid (CS) will be manufactured by assembling a stack of six modules, which are under fabrication by the US ITER organization and its subcontractors. The tests of the first CS Module have been performed at the premises of the General Atomics (GA) facility in Poway (US), in order to check compliance to the ITER requirements. Among other tests, the magnet was submitted to exponential dumps of the transport current from different initial values (10, 15, 20, 22.5, 25, 35, 40 kA) down to 0 kA. These tests are aimed at conducting DC breaker commissioning of the test facility and were used to measure the AC losses in the coil during electrodynamic transients. This paper presents the results of these measurements, along with a comparison with analytical computations of the losses in the magnet.

The Divertor Tokamak Test (DTT) facility, a fully superconducting nuclear fusion reactor being built in Italy, will contribute to address the power exhaust problem in EU DEMO perspective. A lot of flexibility of operation will be demanded to the machine, which should be capable to tackle also severe transients such as plasma disruptions. In this work, the 4C thermal-hydraulic code is used to compute the temperature margin during a plasma disruption, using as input the heat generated into the Toroidal Field coil casing and transferred to the winding pack, and the possibility that this leads to a quench of the magnet is studied. The results of the analysis will give important feedbacks for the design of the quench protection system, e.g., avoiding to trigger a fast current discharge right after the disruption, as well as for the machine operation, e.g., assessing the required re-cooling time of the magnets after a disruption.

Experiments on quench propagation in high-current, high field High Temperature Superconducting (HTS) Cable-In-Conduit Conductors (CICCs) for fusion applications are forthcoming. Among the conductor designs to be tested, there is the ENEA slotted core proposal. In order to support the design of the samples and plan the diagnostics to be employed, a 1D multi-region thermal-hydraulic and electric model of the samples has been developed using the H4C code. After an experimental electric characterization, the model is applied to the simulation of quench propagation in the samples. The simulations show how current redistributes among the tapes and the slots. Additionally, they show that the quench protection strategy is suitable to prevent too high hot-spot temperatures.

The ITER Central Solenoid (CS) will be realized by assembling a stack of six modules. Each module is a solenoid consisting of 40 pancakes wound with a Nb ₃ Sn Cable in Conduit Conductor (CICC). The tests of the second module (CSM#2) were performed at the General Atomics (GA) facility in Poway (US). During the test campaign, the CS Module is submitted to dumps of the transport current from different initial values (10, 15, 20, 25, 30, 35, 40 kA) to 0 kA, which allow measuring the AC losses in the coil. In this paper we present the results on AC losses during the dumps from different initial transport currents, as computed with two different methods. The first method is based on the observation that the dumps determine a very fast pressure rise of the supercritical helium embedded in the module, which undergoes an isochoric transformation. This approach is therefore based on the computation of the variation of internal energy of the helium during the pressure rise itself. The second method is based instead on a calorimetric procedure aimed at estimating the enthalpy variation of the supercritical helium due to the thermal power deposited during the current dumps. The main contribution to the losses during these dumps is due to the coupling losses. As for the hysteresis losses in the magnet, the results obtained through slow current cycle tests are also presented.