1,720,982 research outputs found
Modelling Innovative High Temperature Superconductors for Fusion Applications
Get PDFL'abstract è presente nell'allegato / the abstract is in the attachmen
3D transient CFD simulation of an in-vessel Loss-Of-Coolant Accident in the EU DEMO fusion reactor
Get PDFAccidental transients in fusion reactors, such as the in-vessel Loss-of-Coolant Accident (LOCA) considered here, are generally analyzed using system-level codes. However, because of their lumped nature, such tools cannot predict local pressure and temperature values, which are instead of interest to assess the integrity of the Vacuum Vessel (VV) structures. It is then fundamental to prove that the system-level tools' predictions are at least conservative. In this work, we analyze the helium flow inside the VV following a LOCA in the helium-cooled blanket of the EU DEMO, using a 3D transient CFD model implemented in the commercial STAR-CCM+ code. In view of the large pressure ratio, a hypersonic flow regime (Ma > 5) develops, with the formation of shock fronts requiring a high time and space resolution. The model is applied to compute the evolution of the pressure distribution inside the VV and then to compare it with the information provided by the system-level GETTHEM model. The predictions by the two models of the intervention time of the VV pressure suppression system are compared, showing that the 0D model is conservative
CFD analysis of natural convection cooling of the in-vessel components during a shutdown of the EU DEMO fusion reactor
Get PDFIn view of the large neutron fluence expected in a fusion power plant, the maintenance of the in-vessel components (IVC) must be carried out using Remote Handling (RH); however, before the RH robots can intervene, the temperature of the IVCs must be reduced, so a cooldown phase is required after the reactor shutdown before maintenance activities can start. In the EU DEMO two options are being investigated to cool down the Breeding Blanket (BB) structures before maintenance, namely introducing fans to pump air in forced convection in the plasma chamber (after opening the Vacuum Vessel), or letting the air at room temperature cool down the structures by natural convection; if the required downtime is acceptable, the second option is clearly preferred, as it would reduce the cost and complexity of the system. This work analyses the natural convection option via a 3D transient Computational Fluid-Dynamics (CFD) conjugate heat transfer model, to evaluate the required time to cool down the BB
3D-FOX—A 3D Transient Electromagnetic Code for Eddy Currents Computation in Superconducting Magnet Structures: DTT TF Fast Current Discharge Analysis
Get PDFPredictive 1-D thermal-hydraulic analysis of the prototype HTS current leads for the ITER correction coils
No full textWe present an analysis of the prototype high-temperature superconducting (HTS) current leads (CLs) for the ITER correction coils, which will operate at 10 kA. A copper heat exchanger (HX) of the meander-flow type is included in the CL design and covers the temperature range between room temperature and 65 K, whereas the HTS module, where Bi-2223 stacked tapes are positioned on the outer surface of a stainless steel hollow cylindrical support, covers the temperature range between 65 K and 4.5 K. The HX is cooled by gaseous helium entering at 50 K, whereas the HTS module is cooled by conduction from the cold end of the CL. We use the CURLEAD code, developed some years ago and now supplemented by a new set of correlations for the helium friction factor and heat transfer coefficient in the HX, recently derived using Computational Fluid Dynamics. Our analysis is aimed first of all at a ‘‘blind" design-like prediction of the CL performance, for both steady state and pulsed operation. In particular, the helium mass flow rate needed to guarantee the target temperature at the HX-HTS interface, the temperature profile, and the pressure drop across the HX will be computed. The predictive capabilities of the CURLEAD model are then assessed by comparison of the simulation results with experimental data obtained in the test of the prototype correction coil CLs at ASIPP, whose results were considered only after the simulations were performed
3D Transient CFD Simulation of an In-Vessel Loss-of-Coolant Accident in the EU DEMO WCLL Breeding Blanket
Get PDFThe in-vessel Loss-of-Coolant Accident (LOCA) is one of the design basis accidents in the design of the EU DEMO tokamak fusion reactor. System-level codes are typically employed to analyse the evolution of these transients. However, being based on a lumped approach, they are unable to quantify localised quantities of interest, such as local pressure peaks on the vacuum vessel walls, to which the failure criteria are linked. To calculate local quantities, the 3D nature of the phenomenon needs to be considered. In this work, a 3D transient model of the in-vessel LOCA from a water-cooled blanket is developed. The model is implemented in the commercial CFD software STAR-CCM+. It simulates the propagation of the water jet in the vessel from the beginning of the accident, thus accounting for the phase change of the water, i.e., from the pressurised liquid phase to the vapour phase inside the vessel, being the latter at a much lower pressure than in the blanket coolant pipes. Due to the large pressure ratio (>1000), shocks are expected; therefore, an Adaptive Mesh Refinement (AMR) algorithm is employed. The physical models (in particular, the multiphase model) are benchmarked to a 2D reference problem before being applied to the 3D EU DEMO-relevant problem. The simulation results show that the pressure peaks in front of the vessel walls are not dangerous as they are below the design limit. The entire evolution of the water jet is followed up to the opening of the burst disks, in order to compare the average pressure evolution with that computed with system-level codes. A comparison with the in-vessel LOCA from a helium-cooled blanket is also carried out, showing that the accident evolution in the water case is less violent than in the helium case
Impact of Hysteresis Losses in Hybrid (HTS-LTS) Coils for Fusion Applications
Get PDFSeveral conductor designs featuring High Temperature Superconducting (HTS) stacked tapes for fusion coils are being proposed. These conductors are planned to operate in time-varying magnetic field and current; thus, the estimation of AC losses is fundamental for the conductor design and the accurate analysis of its performance in operation. The case study of an HTS conductor proposed for the hybrid (HTS-LTS) Central Solenoid coil for the EU DEMO tokamak is considered in this work. Here, a numerical model based on the finite element method (FEM) and the H-formulation is used, in order to estimate the hysteresis losses. The FEM model is first benchmarked against available analytical formulae as well as available literature data. Then it is applied to the real case operational scenario. It is shown that for the conductor design analyzed, the coupling losses are orders of magnitude lower than the hysteresis ones. The impact of the hysteresis+coupling losses on the temperature margin of the coil is assessed with a thermal-hydraulic model. It is shown that the heat generated in the HTS layers is partially transferred to the LTS layers, leading these layers to quench. An alternative conductor concept is also analyzed, showing that, however, in the top and bottom modules of the CS coil, due to the bending of the magnetic field, a too large heat deposition is present
Validation of the 4C Code on the AC Loss Tests of a Full-Scale ITER Coil
Get PDFThe AC loss tests on the first ITER Central Solenoid Module (CSM) have been modelled and compared to the test results. The model has been implemented in the 4C code, a thermal-hydraulic modelling tool which includes the CSM winding pack and the cryogenic circuit of the test facility. Two modes of operation of the circuit have been analyzed: the nominal and the “isolation” mode, i.e., when the cryogenic circuit valves are operated to isolate the coil during the current dumps. The computed mass flow rate, pressure and coil outlet temperature at different locations have been compared with the measurements, showing a very good agreement in both modes of operation of the circuit. The validated model helped in the interpretation of the experimental results, such as the backflow at the coil inlet -which cannot be measured- or the non-monotonic outlet temperature evolution following the current dump. Furthermore, the code was used to qualify the isochoric method for the quantification of the deposited energy due to AC losses, as it was the only method applicable in case of current dumps from high current
Thermal-hydraulic analysis of the DTT Toroidal Field magnets in DC operation
Get PDFThe Divertor Tokamak Test (DTT) facility is currently under design in Italy. This fully superconductive compact tokamak will be the test bench of several DEMO-relevant divertor solutions. The 4C code model of a DTT toroidal field magnet (including its structures) is developed here and used to support some important design decisions related to the neutron shield to be adopted, the need of an active cooling of the casing and the static heat load reduction. The simulations confirm the need to actively cool the casing by suitable cooling channels, but also the need of a proper neutron shield to reduce the nuclear heat load on the superconducting coils. On the other hand, the proposed static heat load reduction measures do not appear to be effective enough to satisfy the design require-ment of 1.4 K for the minimum temperature margin
- …
