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Artificial Neural Networks (ANNs) are applied to the development of a simplified transient model of the ITER Central Solenoid (CS), aiming at predicting the evolution of the pulsed heat load from the CS to the LHe bath during plasma operation. The ANNs are trained using the thermal–hydraulic evolution in the CS, computed with the 4C code, due to AC losses. The capability of the ANN model to predict the heat load to the LHe bath is successfully demonstrated in the case of different transients, among which a nominal plasma operating scenario. The gain in speed of the simplified model with respect to the 4C code results is by order of magnitudes, with a small loss of accuracy. [ABSTRACT FROM AUTHOR]

Abstract: The 3D steady-state Computational Fluid Dynamics (CFD) analysis of the ITER vacuum vessel (VV) regular sector #5 is presented, starting from the CATIA models and using a suite of tools from the commercial software ANSYS FLUENT®. The peculiarity of the problem is linked to the wide range of spatial scales involved in the analysis, from the millimeter-size gaps between in-wall shielding (IWS) plates to the more than 10m height of the VV itself. After performing several simplifications in the geometrical details, a computational mesh with ∼50 million cells is generated and used to compute the steady-state pressure and flow fields from a Reynolds-Averaged Navier–Stokes model with SST k-ω turbulence closure. The coolant mass flow rate turns out to be distributed 10% through the inboard and the remaining 90% through the outboard. The toroidal and poloidal ribs present in the VV structure constitute significant barriers for the flow, giving rise to large recirculation regions. The pressure drop is mainly localized in the inlet and outlet piping. [Copyright &y& Elsevier]

Highlights: [•] Innovative approach for the simplified dynamic model of heat load to cryoplants. [•] Artificial Neural Networks (ANNs) developed and applied to HELIOS test loop. [•] ANNs are proved to be accurate, flexible and fast in execution. [•] The novel approach is promising for application to ITER. [ABSTRACT FROM AUTHOR]

Abstract: A new tool – the 4C code – has been developed, which allows the thermal-hydraulic simulation of the entire superconducting magnet system of the International Thermonuclear Experimental Reactor (ITER), with particular reference to: (1) the winding made of cable-in-conduit conductors (CICC), (2) the structures (the radial plates and the case of the toroidal field – TF – coils, for instance) and (3) the cooling circuits. In this paper the different components of the 4C code (1D 2-channel model of the CICC and of the structure cooling channels, 2D model of selected cross sections of the structures, 0D/1D model of the cryogenic circuit) are described in detail, together with the strategy adopted for the coupling between the different components and their integration in a single tool. The new tool is then applied to the modeling of two transients in an ITER TF coil: a simplified version of a cooldown of the coil and the response to a heat pulse applied in the winding. [Copyright &y& Elsevier]

Four Nb3Sn conductor lengths were prepared according to the ITER TF conductor design and assembled into two SULTAN samples. The four lengths are not fully identical, with variations of the strand supplier, void fraction and twist pitch. Lower void fractions improve the strand support and increased twist pitches also lower the strand contact pressure but both tend to increase the AC loss and the lower void fraction also increases the pressure drop so that the mass flow rate in the strand bundle area of the cable is reduced. The assembly procedure of the two samples is described including the destructive investigation on a short conductor section to assess a possible perturbation of the cable-to-jacket slippage during the termination preparation. Based on the DC performance and AC loss results from the test in SULTAN, the impact of the void fraction and twist pitch variations is discussed in view of freezing the ITER conductor design and large series manufacture. A comparison with the former generation of conductors, using similar strands but based on the ITER Model Coil layout, is also carried out. The ITER specifications, in terms of current sharing temperature, are fulfilled by both samples, with outstanding results for the conductor with longer twist pitches. [ABSTRACT FROM AUTHOR]

The Poloidal Field Conductor Insert (PFCI) of the International Thermonuclear Experimental Reactor (ITER) has been designed in the EU and is being manufactured at Tesla Engineering, UK, in the frame of a Task Agreement with the ITER International Team. Completion of the PFCI is expected at the beginning of 2005. Then, the coil shall be shipped to JAERI Naka, Japan, and inserted into the bore of the ITER Central Solenoid Model Coil, where it should be tested in 2005 to 2006. The PFCI consists of a NbTi dual-channel conductor, almost identical to the ITER PF1 and PF6 design, ∼45 m long, with a 50 mm thick square stainless steel jacket, wound in a single-layer solenoid. It should carry up to 50 kA in a field of ∼6 T, and it will be cooled by supercritical He at ∼4.5 K and ∼0.6 MPa. An intermediate joint, representative of the ITER PF joints and located at relatively high field, will be an important new item in the test configuration with respect to the previous ITER Insert Coils. The PFCI will be fully instrumented with inductive and resistive heaters, as well as with voltage taps, Hall probes, pickup coils, temperature sensors, pressure gauges, strain and displacement sensors. The test program will be aimed at DC and pulsed performance assessment of conductor and intermediate joint, AC loss measurement, stability and quench propagation, thermal-hydraulic characterization. Here we give an overview of the preparatory work toward the test, including a review of the coil manufacturing and of the available instrumentation, a discussion of the most likely test program items, and a presentation of the supporting modeling and characterization work performed so far. [ABSTRACT FROM AUTHOR]

Abstract: The computational fluid dynamics (CFD) analysis of the FW06 panel of the ITER shielding blanket is presented in two companion papers. In this Part II we concentrate on the thermal-hydraulics of the water coolant, driven by the nuclear volumetric and plasma surface heat loads discussed in Part I. Both the detailed steady state analysis of a single cooling channel and the coarse transient analysis of the whole panel are considered. The compatibility of the hot spots with the maximum recommended temperatures for the different materials is confirmed. The heat transfer coefficient between coolant and walls is obtained post-processing the results of the simulation and compared with the results of available correlations, which may be used for simpler analyses: in the fully developed flow regions of the cooling pipes, it turns out to be well approximated by the Sieder–Tate correlation. The operation margin with respect to the critical heat flux is also computed and turns out to be sufficiently large compared with the design limit. [Copyright &y& Elsevier]

Highlights: [•] The cryogenic circuit module of 4C is extended to include proportional-integral controllers. [•] We analyze HELIOS tests of pulsed heat load smoothing by a controlled bypass valve. [•] The simulation results are in good agreement with the measurements, without fitting parameters. [ABSTRACT FROM AUTHOR]

Abstract: We complete the first validation campaign of the Cryogenic Circuit Conductor and Coil (4C) code, focusing on the cryogenic circuit module of 4C, which is based on the component models from the recently developed “Cryogenics” Modelica library. Measured data from the HELIOS facility (HElium Loop for hIgh LOads Smoothing) at CEA Grenoble, France, are used as reference. HELIOS includes a supercritical He loop (cold circulator, pipes equipped with resistive heaters, control and bypass valves, heat exchangers) and a saturated He bath. A repetitive heat pulse test is simulated with 4C. The computed evolution of temperature, pressure and mass flow rate at different circuit locations, both in the loop and in the bath, shows a very good agreement with the measurements. [Copyright &y& Elsevier]

Abstract: The recently developed 4C thermal-hydraulic code, currently under validation, is used to compute the temperature margin of the superconducting NbTi toroidal field coil of the ITER satellite tokamak JT-60SA, for the nominal burn operation of the machine. Repetitive conditions in the simulation are reached after two plasma pulses. For given (computed) thermal load distribution on winding and coil case due to nuclear heating only, helium temperature ∼4.4K at the winding inlet, reference strand, and the most recent conductor and winding layout, the computed minimum margin occurs in pancake #7 at the peak field location and turns out to be ∼0.2–0.3K above the design value of 1.2K. [Copyright &y& Elsevier]