Helium Flow and Temperature Distribution in a Heated Dual-Channel CICC Sample for ITER (original) (raw)
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Helium flow and temperatures in a heated sample of a final ITER TF cable-in-conduit conductor
Journal of Physics: Conference Series, 2010
The quest for a detailed understanding of the thermo-hydraulic behaviour of the helium flow in the dual-channel cable-in-conduit conductor (CICC) for the ITER toroidal-field coils led to a series of experiments in the SULTAN test facility on a dedicated sample made according to the final conductor design. With helium flowing through the conductor as expected during ITER operation, the sample was heated by eddy-current losses induced in the strands by an applied AC magnetic field as well as by strip heaters mounted on the outside of the conductor jacket. Temperature sensors mounted on the jacket surface, in the central channel and at different radii in the sub-cable region showed the longitudinal as well as radial temperature distribution at different mass flow rates and heat loads. Spot heaters in the bundle and the central channel created small heated helium regions, which were detected downstream by a series of temperature sensors. With a time of flight method the helium velocity could thus be determined independently of any flow model. The temperature and velocity distributions in bundle and central channel under different mass-flow and heat load conditions thus led to a detailed picture of the helium flow in the final ITER TF CICCs.
Cryogenics, 1998
Mithrandir 1-a 1-D code for the analysis of thermal-hydraulic transients in cable-inconduit super-conductors (CICC) with cooling channel-has been extended to helium II by including Gorter-Mellink heat transport. The code treats the general case of different thermodynamic properties and flow velocities for the helium in the cable bundle region and that in the cooling channel, which are coupled to conductor and jacket at different temperatures. A detailed validation of the code against a thermal-hydraulic experiment in helium II is presented, showing good agreement, and code convergence is demonstrated numerically. We present results of simulations of the QUench Initiation and Propagation Study (QUIPS) experiment, concentrating mostly on the initial phase of the quench. We show that the simulation is at least qualitatively in agreement with experimental data.
Transverse heat transfer coefficients on a full size dual channel CICC ITER conductor
Cryogenics, 2006
Dual channel Cable-In-Conduit Conductors (CICC) provide low hydraulic resistance and faster central channel circulation, limiting superconductors temperature rise. The Poloidal Field Insert Sample (PFIS) was tested in the SULTAN facility to evaluate the thermal coupling between the CICC channels upon an experimental heat transfer coefficient assessment. Simple assumptions on the flow -homogeneous central and annular temperatures, no jacket conduction, no steel inertia and diffusivity -lead to a one-dimensional thermal model fully solved in its transient response to a Heavyside temperature evolution at the inlet, using a Laplace transformation. Transient temperature step data fitted with the analytical resolution provide heat transfer coefficients as a function of mass flow rate, compared to crude predictions. The transient measurements provided consistent measurements on the full range of mass flow rate in both vertical flow directions, whereas steady state homogenization characteristic length measures pursuing the same goal suffer from annular isothermal assumption. Recommendations are made for the thermohydraulic instrumentation of future conductor samples.
Thermal-Hydraulic Issues in the ITER Toroidal Field Model Coil (TFMC) Test and Analysis
AIP Conference Proceedings, 2004
The International Thermonuclear Experimental Reactor (ITER) Toroidal Field Model Coil (TFMC) was tested in the Toska facility of Forschungszentrum Karlsruhe during 2001 (standalone) and 2002 (in the background magnetic field of the LCT coil). The TFMC is a racetrack coil wound in five double pancakes on stainless steel radial plates using NbsSn dual-channel cable-in-conduit conductor (CICC) with a thin circular SS jacket. The coil was cooled by supercritical helium in forced convection at nominal 4.5 K and 0.5 MPa. Instrumentation, all outside the coil, included voltage taps, pressure and temperature sensors, as well as flow meters. Additionally, differential pressure drop measurement was available on the two pancakes DP 1.1 and DP 1.2, equipped with heaters. Two major thermal-hydraulic issues in the TFMC tests will be addressed here: 1) the pressure drop along heated pancakes and the comparison with friction factor correlations; 2) the quench initiation and propagation. Other thermal-hydraulic issues like heat generation and exchange in joints, radial plates, coil case, or the effects of the resistive heaters on the helium dynamics, have been already addressed elsewhere.
Transverse heat transfer coefficient in the dual channel ITER TF CICCs
Cryogenics, 2011
Two ITER TF dual channel Cable-in-Conduit Conductors (CICCs) have been tested in the SULTAN test facility. The samples were heated either by foil heaters mounted on the outside of the conductor jacket or by induced AC losses. The steady-state temperature response of several thermometers installed on the jacket surface as well as inside the cable were analyzed using the two-channel analytical model proposed by Renard et al. to obtain the equivalent transverse heat transfer coefficient between the bundle and central channel as a function of the mass flow rate. In addition, on the basis of the measured pressure drop and helium flow velocities, the friction factors for helium flow in the bundle and in the central channel were determined. The obtained results may serve as a reference for these cables.
The PF-FSJS is a full-size joint sample, based on the NbTi dual-channel cable-inconduit conductor (CICC) design currently foreseen for the International Thermonuclear Experimental Reactor (ITER) Poloidal Field coil system. It was tested during the summer of 2002 in the Sultan facility of CRPP at a background peak magnetic field of typically 6 T. It includes about 3 m of two jointed conductor sections, using different strands but with identical layout. The sample was cooled by supercritical helium at nominal 4.5-5.0 K and 0.9-1.0 MPa, in forced convection from the top to the bottom of the vertical configuration. A pulsed coil was used to test AC losses in the two legs resulting, above a certain input power threshold, in bundle helium backflow from the heated region. Here we study the thermal-hydraulics of the phenomenon with the M&M code, with particular emphasis on the effects of buoyancy on the helium dynamics, as well as on the thermal-hydraulic coupling between the wrapped bundles of strands in the annular cable region and the central cooling channel. Both issues are ITER relevant, as they affect the more general question of the heat removal capability of the helium in this type of conductors.
Fusion Engineering and Design, 2007
We present a new multi-solid multi-channel (M 3 ) thermal-hydraulic model for the analysis of the International Thermonuclear Experimental Reactor (ITER) Cable-In-Conduit Conductors (CICC). The model discretizes the cross section of an ITER CICC into M current carrying cable elements (e.g., the six last-but-one cabling stages-the petals), coupled with N hydraulic channels (e.g., the six petals + the central channel) and K non-current carrying solid components (e.g., the jacket of the CICC), with M, N and K arbitrary integers. Along each of the M + K solid components a 1D transient heat conduction equation is solved, whereas along each of the N channels three Euler-like 1D equations, derived from the conservation laws for compressible He flow, are solved. The resulting quasi 3D model, in which 1D equations are coupled by heat and mass transfer between the different CICC components, is implemented in the M 3 code and validated against experimental results from the ITER Good Joint sample and the ITER Poloidal Field Conductor Insert Full Size Joint Sample. The new code is able to reproduce with good accuracy the measured temperature gradients on the CICC cross section, provided sufficiently accurate input data are available.
Evaluation of thermal gradients and thermosiphon in dual channel cable-in-conduit conductors
Cryogenics, 2006
In an effort to optimize superconductor cryogenics of large coils, dual channel cable-in-conduit conductors (CICC) have been designed. The qualitative and economic rationale of the conductor central channel is here justified but brings high complexity to the conductor cooling characteristics. Temperature gradients in the cable must be quantified to guarantee conductor temperature margin during coil operation under heat disturbance and set adequate inlet temperature. A simple one-dimensional thermal model, with neither fluid nor strand or jacket conduction, allows to better understand and quantify the steady state behavior of CICC central and annular channels. This thermohydraulic model with homogeneous central and annular temperatures and no jacket conduction is summarized with explicit thermal coupling equations. Local convection coefficients chosen proportional to friction factors lead to a model of global interchannel heat exchange coefficient serving the bithermal model. A first stationary experimental evaluation of the internal heat transfer coefficient using the interchannel heat exchange space constant at various heat loads and mass flow rates is illustrated on two full size samples tested at cryogenic temperatures. Annular heaters experiments with low distributed power achieve pertinent model correlation. Discrepancy between model and experimental data may be linked to the simplistic homogeneous annular temperature hypothesis, to the estimate of CICC mass flow distribution among channels, and to gravitational effects at high heat loads. Perturbation due to the thermosiphon generated between the two channels is considered since neither the experiments nor the expected applications are free of gravity.
Two-fluid analysis of the thermal-hydraulic stability of ITER CS and TF superconductors
Thermal-hydraulic stability of the two-channel cable-in-conduit conductors (CICC) foreseen for the central solenoid (CS) and the toroidal field (TF) Nb3Sn super-conducting coils of the International Thermonuclear Experimental Reactor (ITER) is analyzed with the two-fluid code MITHRANDIR [1]. In all externalheating scenarios considered, the computed stability margin is of the order of some 100 mJ/ccst. Extra-strands Cu addition typically leads to higher computed margins. MITHRANDIR estimates are typically conservative with respect to onefluid results.