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.

Bulky metallic structures are needed in the toroidal field (TF) superconducting magnets for fusion applications to withstand the large Lorentz forces acting on the winding. The pulsed coil operation during a plasma scenario and the fast current discharge, or a plasma disruption event in off-normal operating conditions, cause transient magnetic fields, inducing eddy currents in the TF structures. The eddy currents generate heat in the structures heating in turn the winding pack, eroding the temperature margin: such power deposition is a key input for thermal-hydraulic (TH) analyses. However, the computation of eddy currents in fusion magnets is a challenging topic since a transient, fully 3D electromagnetic (EM) model is required. The EM problem is solved here by means of the finite element (FE) open source code FreeFEM++. First, the correct implementation of the EM problem is verified by means of suitable benchmarks against both simple analytical cases and the results obtained with state-of-the-art FE commercial codes on the DTT TF coil, used as a reference geometry. Then the EM code is applied to the evaluation of the magnetic fields and eddy currents induced in the same reference coil during the normal (static) and off-normal (transient) operation; the output of the EM analysis is used as input to the TH analysis carried out with the 4C code, aimed at computing the temperature margin evolution during the transient.

Nuclear fusion reactors are composed of several complex components whose behavior may be not certain a priori. This uncertainty may have a significant impact on the evolution of fault transients in the machine, causing unexpected damage to its components. For this reason, a suitable method for the uncertainty propagation during those transients is required. The Monte Carlo method would be the reference option, but it is, in most of the cases, not applicable due to the large number of required, repeated simulations. In this context, the Polynomial Chaos Expansion has been considered as a valuable alternative. It allows us to create a surrogate model of the original one in terms of orthogonal polynomials. Then, the uncertainty quantification is performed repeatedly, relying on this much simpler and faster model. Using the fast current discharge in the Divertor Tokamak Test Toroidal Field (DTT TF) coils as a reference scenario, the following method has been applied: the uncertainty on the parameters of the Fast Discharge Unit (FDU) varistor disks is propagated to the simulated electrical and electromagnetic relevant effects. Eventually, two worst-case scenarios are analyzed from a thermal–hydraulic point of view with the 4C code, simulating a fast current discharge as a consequence of a coil quench. It has been demonstrated that the uncertainty on the inputs (varistor parameters) strongly propagates, leading to a wide range of possible scenarios in the case of accidental transients. This result underlines the necessity of taking into account and propagating all possible uncertainties in the design of a fusion reactor according to the Best Estimate Plus Uncertainty approach. The uncertainty propagation from input data to electrical, electromagnetic, and thermal hydraulic results, using surrogate models, is the first of its kind in the field of the modeling of superconducting magnets for nuclear fusion applications.

In the last years several experimental campaigns have been carried out to assess the performance of High Temperature Superconducting (HTS) cables for nuclear fusion applications. In such experiments it has been often observed that cable performance in terms of critical current $(I_{c})$ degrade, mainly after quenches. The nature of the degradation is not completely assessed, but secondary strains given by temperature increase are considered to be its responsible. To confirm such statement it is necessary to develop a tool capable of reproducing the quench effects both from the thermal and from the mechanical points of view, aiming at retrieving the induced strain field in the conductor itself. The aim of this work is presenting the development of a simulation environment which embeds both the required features mentioned above. The thermal-hydraulic (TH) aspects of the quench are analysed here with the H4C code, developed at Politecnico di Torino and spreadly validated against several quench campaigns in SULTAN. The output of the H4C code represents an optimum set of inputs for a detailed mechanical model of the cable. For this reason a python API has been developed making the H4C code communicate with ANSYS, used to compute, on the base of the temperature profiles in input, the consequent strain field. The simulation environment developed has been applied to a slotted core HTS conductor to check its functionalities, showing good agreement with the physical expectations.

Nuclear fusion reactors are composed by several complex components whose behaviour may be not certain a priori. This uncertainty may have a significant importance in the evolution of fault transients in the machine, causing unexpected damages to its components. For this reason a suitable method for the uncertainty propagation during those transient is required. The Monte Carlo method would be the reference option, but it is, in most of the cases, not applicable due to the large amount of required simulations to be repeated. In this context the Polynomial Chaos Expansion has been considered as a valuable alternative. It allows to create a surrogate model of the original one in terms of orthogonal polynomials. Then, the uncertainty quantification is performed repeatedly calling this much simpler and faster model. Using the fast current discharge in the Divertor Tokamak Test Toroidal Field (DTT TF) coils as a reference scenario, this method has been applied: the uncertainty on the parameters of the Fast Discharge Unit (FDU) varistor disks is propagated to the simulated electrical and electro-magnetic relevant effects. Eventually two worst case scenarios have been analyzed from the thermal-hydraulic point of view with the 4C code, simulating a fast current discharge as a consequence of a coil quench. Eventually it has been demonstrated that the uncertainty on the inputs (varistor parameters) strongly propagates obtaining a wide range of possible scenarios in case of accidental transients. This result underlines the necessity to take into account and propagate all possible uncertainties in the design of a fusion reactor, according to the Best Estimate Plus Uncertainty approach.