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The recently developed and validated 4C code is used to analyse the fast discharge of an ITER TF coil, including the cooling circuits of winding and casing, and the quench venting system. The analysis shows that a quench can be initiated by the fast discharge, depending on the value of the thermal resistance of the conductor turn insulation. Eddy currents heat the structures and these transfer the heat especially to the side pancakes, where the quench is initiated. The hot spot temperature remains below \sim37 K in the winding and \sim55 K in the structures. The peak pressurization in the winding remains below \sim8.5 MPa. [ABSTRACT FROM PUBLISHER]

We apply the 4C code, recently validated against different types of transients, to the simulation of the first stage of the cool-down of an ITER Toroidal Field (TF) magnet, from 300 K to 80 K. For a given scenario (inlet temperature, pressure and total mass flow rate evolution), we compute the evolution of the maximum temperature difference across the magnet. These results are compared both with the allowable and with the previously published results of another code, thus providing a benchmark for 4C. The differences between the models implemented in the two codes, whenever present, are highlighted, and their effects are analyzed by suitable sensitivity studies. [ABSTRACT FROM PUBLISHER]

The toroidal field model coil (TFMC) was extensively tested in 2001–2002, in the TOSKA facility at the Karlsruhe Institute of Technology, Germany, up to a maximum transport current of 80 kA and a peak magnetic field of \sim10 T. Here, we apply the 4C code, which was developed for the analysis of thermal-hydraulic transients in superconducting magnets, including winding, structures, and cryogenic circuit, to the analysis of the TFMC safety discharge from 80 kA. The code is able to reproduce the main experimental features, despite the limitations due to no more fully available information about the details of the cryogenic circuit, which play here a much more important role than at lower currents. [ABSTRACT FROM PUBLISHER]

We apply a recently proposed Computational thermal Fluid Dynamics (CtFD) strategy to the analysis of a meander-flow path (MF) fin-type heat exchanger (HX), to be used in the HTS current leads for the LTS coils of both the W7-X stellarator and the JT-60SA tokamak. A mock-up of the HX was tested at the Karlsruhe Institute of Technology providing the database for the validation of the computational model. The hydraulic characterization of the mock-up is considered first, and then the heat transfer characteristic is analyzed. [ABSTRACT FROM AUTHOR]

The 1D stability analysis of a subset of the International Thermonuclear Experimental Reactor (ITER) Poloidal Field (PF) coils, namely PF2, PF3, PFS, is presented, using the Mithrandir code. Different combinations of lengths and durations are considered for the stability disturbance, applied to the superconducting cable during the most critical operating condition. In all cases the computed energy margin is well above the presently expected disturbance on the coils. [ABSTRACT FROM AUTHOR]

Pressure drop in the ITER PFCI cable-in-conduit conductor (CICC) has been measured at CRPP using pressurized water at room temperature. The PFCI conductor is a dual channel CICC and the coolant flows in parallel in the central channel and in the annular bundle region. In our experiment the flow in the central channel is blocked and the longitudinal friction factor of the annular bundle region is deduced from measurements of pressure drop and mass flow rate. Two conductor samples are investigated, one with and one without subcable/outer cable wraps. The results show that the wraps have a negligible effect on the friction factor, and that the Katheder correlation overestimates the actual friction factor. [ABSTRACT FROM AUTHOR]

Abstract: The Karlsruhe Institute of Technology (KIT) is responsible for the design, construction and testing of the high temperature superconductor current leads for the Wendelstein 7-X (W7-X) stellarator. Two prototypes of the W7-X current lead, using a meander-flow heat exchanger, have been tested in 2010 at KIT. Following an approach originally developed and applied to a sub-size mock-up heat exchanger, the Computational Fluid Dynamics code StarCD is used here for the 3D thermal-hydraulic analysis of the heat exchanger in the prototypes. The resulting friction and Nusselt parameters are then fed to a global 1D model of the current lead and the results compared to the measured pressure drop and temperature profiles along the heat exchanger. [Copyright &y& Elsevier]

• Analysis of the cooldown of an EU DEMO sector. • 3D transient CFD analysis of the Conjugate Heat Transfer problem. • Detailed model including half of the Plasma Chamber. • Timing needed to allow intervention of Remote Handling identified. In 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. [ABSTRACT FROM AUTHOR]

The ITER Central Solenoid (CS) is under fabrication by the US ITER organization and its subcontractors. US ITER will supply seven modules to ITER IO, six of which will be assembled in a stack that forms the ITER Central Solenoid, with one as a spare. The first module fabrication has been completed by General Atomics (GA) at their facility and has begun testing including high voltage testing, Paschen testing in the vacuum and then testing at 4.5 K and up to 40 kA in order to demonstrate compliance of the coil to ITER requirements. In the paper we present the Test Plan and results of the CS Module performance tests, especially at 40 kA current. AC losses, joint resistances and hydraulic characteristics of the coil are all measured. Displacements of the coil height and hoop strain of the CS Module are also measured to verify structural and mechanical characteristics of the coil along with cooldown shrinkage of the coil. This information is used for verification of the stack behavior of CS in ITER operation. The test results and preliminary analyses results are presented, compared to expectations, and discussed. All measured parameters suggest that the CS module will perform well in ITER machine. [ABSTRACT FROM AUTHOR]