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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.

The 3D Computational Fluid Dynamic (CFD) steady state analysis of the regular sector #5 of the ITER vacuum vessel (VV) is presented in these two companion papers using the commercial software ANSYS-FLUENT®. The pure hydraulic analysis, concentrating on flow field and pressure drop, is presented in Part I. This Part II focuses on the thermal-hydraulic analysis of the effects of the nuclear heat load. Being the VV classified as safety important component, an accurate thermal-hydraulic analysis is mandatory to assess the capability of the water coolant to adequately remove the nuclear heat load on the VV. Based on the recent re-evaluation of the nuclear heat load, the steady state conjugate heat transfer problem is solved in both the solid and fluid domains. Hot spots turn out to be located on the surface of the inter-modular keys and blanket support housings, with the computed peak temperature in the sector reaching ∼290 °C. The computed temperature of the wetted surfaces is well below the coolant saturation temperature and the temperature increase of the water coolant at the outlet of the sector is of only a few °C. In the high nuclear heat load regions the computed heat transfer coefficient typically stays above the 500 W/m2 K target.

The pulsed heat load to the cryoplant is an important issue in the design and operation of tokamaks adopting superconducting (SC) magnets for the magnetic confinement, as the International Thermonuclear Experimental Reactor (ITER). The smoothing of the heat load during plasma operation is being addressed by experiments, e.g. in the HELIOS facility at CEA Grenoble, and simulations. The assessment of the operation of the cryoplant mainly requires the knowledge of the evolution of the heat load to the liquid helium (LHe) baths that are used as interfaces/buffers between the magnets cooling loops and the cryoplant itself. In this paper, an innovative approach based on Artificial Neural Networks (ANNs) is presented, leading to a simplified but fast model of the transient heat load from the magnets to the LHe baths. An ANN model is developed for the HELIOS loop and the resulting network is trained using detailed transient simulations performed with the 4C code, which was previously extensively validated against experimental data from HELIOS. The predictive capability of the (simplified) ANN model is then demonstrated by considering another, independent dataset, not used during the ANN training, and comparing the evolution of the heat load to the LHe bath computed by the ANNs with that obtained from the (detailed) 4C model.

Large variations of the heat load coming from the superconducting magnets to the cryoplant in a tokamak are an undesired consequence of the intrinsically pulsed operation of the machine. In this paper, the cryogenic circuit module of the 4C (Cryogenic Circuit Conductor and Coil) code is used to analyze a subset of the results on pulsed load driven transients experimentally simulated in 2011 in the HELIOS facility (HElium Loop for hIgh lOads Smoothing) at CEA Grenoble, France. We focus on the tests where the smoothing of the pulsed heat load from the supercritical helium loop to the saturated helium bath was achieved by means of a controlled bypass valve. To address this issue, the modeling capability of the cryogenic circuit module of the 4C code is extended to include PI controllers. The computed evolution of temperature, pressure and mass flow rate at different circuit locations shows a good agreement with the measurements, both in the SHe loop and in the LHe bath. These results confirm the accuracy of 4C and contribute a needed step in the process of validation of the code.

A new multi-physics simulation tool – FRENETIC (Fast REactor NEutronics/Thermal-hydraulICs) – is presented for the quasi-3D analysis of a lead-cooled fast reactor core with the hexagonal fuel element configuration, as currently proposed within the framework of the European project LEADER. The tool implements coupled neutronic (NE) and thermal-hydraulic (TH) models. In the NE module, a 2D + 1D full-core multi-group diffusion solver has been developed based on a coarse-mesh nodal scheme and adapted to cope with the hexagonal geometry. In the TH module, the hexagonal elements, described by 1D (axial) transient advection and conduction in the coolant coupled to conduction in the fuel pins, are thermally coupled to each other in the transverse directions to obtain the full-core evolution of the distribution of the TH variables. The NE and TH modules are coupled at each TH time step by transferring to the TH module the distribution of the power source computed by the NE module, which is the driver of the TH evolution; alternately, the temperature distribution computed by the TH module is input to the NE module, in order to update the cross sections. The code is benchmarked against pure TH and pure NE analytical solutions and the results of a coupled NE/TH pseudo-transient (criticality search) are also presented. The convergence of the numerical solution is demonstrated both in space and time by computational simulations.

KSTAR is a fully superconducting tokamak, in operation since 2008 at the National Fusion Research Institute in Korea. All coils are wound using cable-in-conduit conductors and cooled with forced-flow supercritical helium (SHe) at 4.5 K and 5.5 bar. We consider here the central pair, PF1U/L, of the central solenoid coils; during operation these coils are subjected to sharp current transients, which induce AC losses in the coil. The thermal hydraulic transient following a trapezoidal current pulse, with ramp up to 10 kA at a rate of 1 kA/s and ramp down to 0 kA at a rate of 10 kA/s, is simulated here using the 4C code, and the results are compared with the measurements.

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 ~ 10 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.

The recently developed 4C thermal-hydraulic code is validated here against experimental data from the cool-down of a non-planar coil of the Wendelstein 7-X stellarator from room temperature to the superconducting transition temperature of ∼10 K, performed during the cold test of the full set of coils at the cryomagnetic test facility of CEA Saclay. The computed results, i.e., temperature and mass flow rate evolutions in different coil components (helium, casing, etc.) are shown to be in good agreement with the experimental data. The simulations also show that during the transient the hot spot temperature inside the coil can be up to ∼3–4 K higher than what is seen on the casing surface.

Dual-channel cable-in-conduit conductors (CICC) are used in the superconducting magnets for the International Thermonuclear Experimental Reactor (ITER). As the CICC axial/transverse size ratio is typically ∼1000, 1D axial models are customarily used for the CICC, but they require constitutive relations for the transverse fluxes. A novel approach, based on Computational Fluid Dynamics (CFD), was recently proposed by these authors to understand the complex transverse thermal–hydraulic processes in an ITER CICC from first principles. Multidimensional (2D, 3D) Reynolds-Averaged Navier–Stokes models implemented in the commercial CFD code FLUENT were validated against compact heat exchanger and ITER-relevant experimental data, and applied to compute the friction factor and the heat transfer coefficient in fully turbulent spiral rib-roughened pipes, mimicking the central channel of an ITER CICC. That analysis is extended here to the problem of heat and mass transfer through the perforated spiral separating the central channel from the cable bundle region, by combining the previously developed central channel model with a porous medium model for the cable region. The resulting 2D model is used to analyze several key features of the transport processes occurring between the two regions including the relation between transverse mass transfer and transverse pressure drop, the influence of transverse mass transfer on axial pressure drop, and the heat transfer coefficient between central channel and annular cable bundle region.

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

The crucial multi-physics problem of how to extrapolate from the performance of an isolated Nb3Sn strand measured in the laboratory to the performance of a superconducting coil using multi-strand twisted cables is addressed here. We consider the particular case of the path going from the LMI strand to the international thermonuclear experimental reactor (ITER) toroidal field model coil (TFMC), through its associated Full Size Joint Sample, the TFMC-FSJS. Mechanical, electromagnetic and thermal–hydraulic conditions are simulated using the ANSYS, ENSIC and Mithrandir/M&M codes, respectively. At least in this case, the DC performance of the short sample turns out to be relatively close to (considering error bars) but not fully representative of that of the coil, showing higher (less compressive) effective thermal strain but also higher sensitivity to the electromechanical load.

The coupled thermal–hydraulic electromagnetic model, implemented in the THELMA code, is applied to the analysis of the sudden quench in the ITER NbTi poloidal field conductor insert full-size joint sample. The computed results are compared with the measured values, showing that the major experimental features, like the premature and sudden voltage take-off, as well as the presence of precursors of the quench in the form of voltage spikes, can be qualitatively reproduced by the code. A self-consistent explanation of the phenomenon is presented, emphasising the effect of current, magnetic field and temperature non-uniformities on the cable cross section, together with the fundamental role played by coupled electromagnetic and thermal–hydraulic dynamics in the current redistribution associated with the voltage spikes.

During 2000, AC losses and the effects of possible ramp-rate limitation (RRL) were investigated on the International Thermonuclear Experimental Reactor (ITER) Central Solenoid Insert Coil (CSIC), at JAERI Naka, Japan. The CSIC was mounted inside the bore of the ITER Central Solenoid Model Coil (CSMC), at the maximum field of about 13 T and experiencing the largest magnetic field variations. The thermal-hydraulic response of the coil to different transport current scenarios was assessed by measuring the temperature increase and pressurization of the supercritical helium (SHe) coolant, together with the evolution of the mass-flow rate. Here we implement in the M&M code a detailed general model of AC losses, which is being validated for the first time. The resulting tool is then applied to the analysis of two CSIC tests, with different ramp-up of the transport current followed by the same dump, and used to qualitatively assess the major thermal-hydraulic effects of AC losses in the coil.

The construction and testing of the Toroidal Field Model Coil (TFMC) is part of one of the ITER large R&D projects. The main goal was to demonstrate the feasibility and the mechanical integrity of the design. One of the highlights of the first test phase was to measure the current sharing temperature, T/sub CS/, of the conductor by heating the helium entering from the inlet. Because neither temperature sensors nor voltage taps are positioned inside the coil, only the helium inlet temperature and the voltage along the whole conductor length can be used for the evaluation of T/sub CS/. In addition, an inner pancake joint is located at the inlet in a rather high magnetic field and the peak field region is only about 1.5 m apart from the joint. The determination of the T/sub CS/ relies on the exact knowledge of the thermohydraulics of both the joint and the conductor region. The paper describes and compares the different numerical models used for the evaluation of the T/sub CS/. Nine T/sub CS/ tests at different coil currents were performed, all ending in a quench. The measured T/sub CS/ is in good agreement with the expectations.

The International Thermonuclear Experimental Reactor Toroidal Field Model Coil, a large (2.7 m × 3.8 m × 0.8 m) superconducting (Nb 3 Sn) DC coil designed and constructed in collaboration between EU industries and laboratories coordinated by European Fusion Development Agreement, has been tested during 2001 in the TOSKA cryogenic facility at Forschungszentrum Karlsruhe, Germany, achieving the nominal 80 kA at 7.8 T peak field and 86 MJ stored energy as a standalone coil (Phase I). Possibly the highlight of the tests was the measurement of the current sharing temperature ( T CS ) at different transport currents I in the self-field of the coil. The measurement method is discussed, based also on the two-year long predictive work, which preceded it. The results of the full set of T CS measurements at I =80, 69 and 57 kA are presented here, and evaluated in a companion paper (part 2).

The ITER Toroidal Field Model Coil (TFMC), a large (2.7 m � 3.8 m � 0.8 m) superconducting (Nb3Sn) DC coil designed and constructed in collaboration between EU industries and laboratories coordinated by EFDA, has been tested during 2001 in the TOSKA cryogenic facility at Forschungszentrum Karlsruhe, Germany, achieving the nominal 80 kA at 7.8 T peak field and 86 MJ stored energy as a standalone coil (Phase I). The results of the current sharing temperature (TCS) measurements at I ¼ 80, 69 and 57 kA, presented in a companion paper (Part 1), are evaluated here using the M&M code. The critical properties best fitting the experimental voltage-inlet temperature characteristic of the P1.2 pancake are deduced from the TFMC data under the assumption of an ideal collective behaviour of the strands. The TFMC results are compared first with the expected conductor performance, showing that at the maximal current the performance was borderline with what was expected, while at the minimal current tested it was better than expected. Second, they are compared with the performance of the single strand as measured in the lab, showing that, in order to reproduce the TFMC data, one has to invoke that some degradation, larger at higher current, occurred when going from the strand to the cable. � 2003 Elsevier Science Ltd. All rights reserved.

The 4C code was developed to model thermal-hydraulics in superconducting magnet systems and related cryogenic circuits. It consists of three coupled modules: a quasi-3D thermal-hydraulic model of the winding; a quasi-3D model of heat conduction in the magnet structures; an object-oriented a-causal model of the cryogenic circuit. In the last couple of years the code and its different modules have undergone a series of validation exercises against experimental data, including also data coming from the supercritical He loop HELIOS at CEA Grenoble. However, all this analysis work was done each time after the experiments had been performed. In this paper a first demonstration is given of the predictive capabilities of the 4C code cryogenic circuit module. To do that, a set of ad-hoc experimental scenarios have been designed, including different heating and control strategies. Simulations with the cryogenic circuit module of 4C have then been performed before the experiment. The comparison presented here between the code predictions and the results of the HELIOS measurements gives the first proof of the excellent predictive capability of the 4C code cryogenic circuit module.

The first detailed Computational Fluid Dynamics (CFD) analysis of the FW06 panel of the ITER shielding blanket is presented in two companion papers. In this Part I we introduce the problem, define the model together with its input and discuss the results with particular reference to the hydraulics of the water coolant. The pressure drop across the panel is computed, together with the distribution of the flow among the different channels. Different design options are studied, with particular reference to the minimization of stagnation/recirculation regions.

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.

The thermal–hydraulic transient, driven in the EAST Central Solenoid by two successive sinusoidal current pulses of different amplitude and frequency, forced in the CS1 coil alone, is studied with the 4C code. The sensitivity of the evolution of the computed transient to the uncertainty of the conductor void fraction and of the Inter-Turn/Inter-Pancake heat transfer is studied parametrically. The coupling time constant (∼36 ms), needed to reproduce with the simulations the measured evolution of the temperature of the supercritical He coolant at the coil outlet, turns out to be much longer than that of the corresponding short sample (∼5 ms).

The forthcoming deuterium-tritium (DT) campaign at the Joint European Torus (JET) will induce a significant activation of the system components. In the present work we evaluate the temporal evolution of the radioactive species in the main in-vessel components after the end of the future DT campaign, assuming different levels of neutron irradiation. The neutron flux in the selected components is calculated by the MCNP5 code using the emission source by a typical DT plasma. The resulting neutron spectra are then input to the FISPACT code that computes the evolution of the radioactive species generated by the neutron activation process. For each irradiation scenario, the time behavior of the contact dose rate is determined. © 2014 Elsevier B.V.

The modeling of the effects of bending on single Nb3Sn strand DC performance (IC, n index) is presented for a bronze-route strand subjected to the same loading conditions as in an experiment performed at JAEA Naka, Japan [Y. Nunoya, et al., IEEE TAS 14 (2004) 1468–1472]. The strand is discretized in strand elements (SE) representing groups of twisted filaments in the bronze matrix, and in portions of the outer Cu annulus, electro-magnetically coupled in the THELMA code. The 3-D strain map in the filament region is computed with a newly developed, detailed thermo-mechanical model accounting for non-linear, temperature dependent material characteristics. With respect to our previous analysis [P.L.Ribani, et al., IEEE TAS 16 (2006) 860–863] several new updated ingredients, besides the new thermo-mechanical model, are used here, including more accurate thermal and mechanical properties for the materials, a jacket-like model for the outer Cu layer, IC and n index (interpolative) scaling from Durham University. The simulation results show an improved agreement with the experiments, in the degradation of the single-strand performance due to bending.

A first validation of the full version of the thermal–hydraulic electromagnetic (THELMA) code, developed for the analysis of transients in the cable-in-conduit conductors and coils relevant for the International Thermonuclear Experimental Reactor (ITER) is presented here. THELMA includes electromagnetic models of the cable joints and terminations (lumped parameter) and of the conductor (distributed parameter), while for the thermal–hydraulics of the helium coolant it includes a compressible 1D flow model. The AC losses induced by a pulsed coil in the NbTi poloidal field full size joint sample (PF-FSJS) right leg conductor, tested in 2002 at the Sultan facility in Villigen (CH), are considered as test bed for this exercise. The computed energy deposition and evolution of the temperature downstream of the heated zone are in good agreement with the measured values. However, the inter-bundle electrical conductances needed in input by the code are compatible with measured values only when a sufficiently refined model is used in the cable cross-section.

The forthcoming Deuterium-Tritium (DT) campaign at the Joint European Torus (JET) will induce a significant activation of the system components. In the present work we evaluate the temporal evolution of the radioactive species in the main in-vessel components after the end of the future DT campaign, assuming different levels of neutron irradiation. The neutron flux in the selected components is calculated by the MCNP5 code using the emission source by a typical DT plasma. The resulting neutron spectra are then input to the FISPACT code that computes the evolution of the radioactive species generated by the neutron activation process. For each irradiation scenario, the time behavior of the contact dose rate is determined.

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 10 m 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.

The KSTAR tokamak, in operation since 2008 at the National Fusion Research Institute in Korea, is equipped with a full superconducting magnet system including the central solenoid (CS), which is made of four symmetric pairs of coils PF1L/U–PF4L/U. Each of the CS coils is pancake wound using Nb 3 Sn cable-in-conduit conductors with a square Incoloy jacket. The coils are cooled with supercritical He in forced circulation at nominal 4.5 K and 5.5 bar inlet conditions. During different test campaigns the measured temperature increase due to AC losses turned out to be higher than expected, which motivates the present study. The 4C code, already validated against and applied to different types of thermal–hydraulic transients in different superconducting coils, is applied here to the thermal–hydraulic analysis of a full set of trapezoidal current pulses in the PF1 coils, with different ramp rates. We find the value of the coupling time constant nτ that best fits, at each current ramp rate, the temperature increase up to the end of the heating at the coil outlet. The agreement between computed results and the whole set of measured data, including temperatures, pressures and mass flow rates, is then shown to be very good both at the inlet and at the outlet of the coil. The nτ values needed to explain the experimental results decrease at increasing current ramp rates, consistently with the results found in the literature.

In the International Thermonuclear Experimental Reactor (ITER), the nuclear radiation escaping from the vacuum vessel reaches the superconducting toroidal field (TF) magnets, contributing to the reduction of the temperature margin ΔTmar, i.e., of the difference between the current sharing temperature and the operating temperature. The TF magnets are designed to operate at a minimum ΔTmar of 0.7 K. However, recent design activity on in-vessel components, e.g., the blanket and the in-vessel coils for plasma stability, suggests a potential enhancement of the nuclear heat load on the TF coils. A detailed reassessment of ΔTmar in the TF magnets during plasma operation is thus required, together with the study of possible strategies to mitigate the corresponding margin reduction. The extensively validated Cryogenic Circuit Conductor and Coil (4C) code, developed for the thermal-hydraulic analysis of transients in superconducting magnets, including their cooling circuit, is used here for the simulation of the standard ITER operating scenario. Among the different possible mitigation strategies, we consider the variation of the cooling He mass flow rate, both in the winding and, separately, in the casing, and its effectiveness in terms of ΔTmar is assessed.

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.

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 ∼4.4 K 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 ∼0.2–0.3 K above the design value of 1.2 K.

The first realistic application of the recently developed 4C code is presented, aimed at showing its capability to simulate in an integrated fashion relevant transients for the superconducting coil operation in the International Thermonuclear Experimental Reactor (ITER), both at the system and at the conductor levels. The quench initiation and propagation in an ITER TF coil is considered, including the coil fast discharge phase until the opening of the relief valves. The 14 coil pancakes cooled by alternating clockwise/counter-clockwise SHe flow, the radial plates and the case, the 96 case cooling channels, and the external cryogenic circuit up to the LHe bath are included in the model. The results of the analysis are discussed with particular reference to quench propagation in the winding, hot spot temperature and peak pressurization, mass flow rate evolution in the different system components. The accuracy of the analysis is guaranteed by suitable convergence studies.

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.

The stability analysis of the International Thermonuclear Experimental Reactor (ITER) Central Solenoid (CS) coil Nb3Sn conductor is performed following a similar approach to that recently used for the ITER Toroidal Field and Poloidal Field conductors. The most critical conductors in the winding pack, as well as the most critical (minimum temperature margin) location along them, are identified by the application of the Vincenta code, which also provides the initial and boundary conditions in the reference case. The Mithrandir code is then applied to these conductors, using a much finer grid than affordable in the Vincenta analysis, in order to capture the details of normal zone initiation and possible recovery to SC state when different square wave disturbances of length L in the range [0.01 m, 7 m] and duration τ the range [1 ms, 100 ms] are applied to the cable. The computed minimum quench energy is shown to be typically above (in one case borderline to) the expected disturbance. The sensitivity to param...

The stability analysis of the reference Nb3Sn conductor for the International Thermonuclear Experimental Reactor (ITER) Toroidal Field (TF) coils is performed using the Mithrandir code. From the point of view of the temperature margin, the most critical conductor in the winding pack, as well as the most critical location along it, is identified by a Vincenta code analysis, which also provides the initial and boundary conditions for the stability study. With this approach, the 1D Mithrandir analysis can be restricted to the most critical conductor, using a much finer grid than Vincenta, in order to capture the details of normal zone initiation and possible recovery to SC state. Two different disturbances are considered: one short in space and time (0.01 m, 1 ms), simulating a disturbance of mechanical nature, the other longer (3 m, 100 ms), corresponding to AC losses (plasma disruption). Both disturbances are applied to the superconducting cable at end-of-burn, in the reference ITER inductive operation sce...

The available database for the friction factors to be used in the annular cable and in the central channel regions of cable‐in‐conduit conductors (CICC) for the superconducting magnets of the International Thermonuclear Experimental Reactor (ITER) is reviewed here for the first time. There is particular reference to its internal consistency as well as the correlations presently used in the ITER design criteria. Limitations and issues in the present status are emphasized and recommendations for improvements are given.

Problems related to cable-in-conduit conductors (CICC) are intrinsically multi-physics involving coupled electro-magnetic/mechanical/thermal-hydraulic fields. Here we concentrate on the thermal-hydraulic issues because, although the CICC was first proposed for the low-TC superconducting coils of the International Thermonuclear Experimental Reactor (ITER) many years ago, CICC thermal-hydraulics alone is less understood than could be expected. Some of the difficulties are due to the multi-channel nature of the ITER CICC, where strands containing the superconducting filaments are twisted in multi-stage sub-bundles (petals) delimited by wrappings and concentrated in an annular (porous-medium like) region, while a central channel, delimited by a spiral, provides lower hydraulic impedance and pressure relief to the flow of the supercritical helium coolant. Other difficulties are related to the multi-scale nature of this problem, with length scales relevant for thermal-hydraulics ranging from the strand diameter (

The coolant in the ITER cable‐in‐conduit conductors (CICC) flows at significant higher speed in the central channel than in the strand bundle region due to the large difference of hydraulic impedance. When energy is deposited in the bundle region, e.g. by ac loss or radiation, the heat removal in vertically oriented dual channel CICC with the coolant flowing downward is affected by the reduced density of helium (buoyancy) in the bundle region, which is arising from the temperature gradient due to poor heat exchange between the two channels. At large deposited power, flow stagnation and back‐flow can cause in the strand bundle area a slow temperature runaway eventually leading to quench.A new test campaign of the thermal‐hydraulic behavior was carried out in the SULTAN facility on an instrumented section of the ITER Poloidal Field Conductor Insert (PFIS). The buoyancy‐gravity effect was investigated using ac loss heating, with ac loss in the cable calibrated in separate runs. The extent of upstream tempera...

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.

The ITER Toroidal Field Model Coil (TFMC) was designed and manufactured by the European Home Team in collaboration with European industry. The test in the TOSKA facility of the Forschungszentrum Karlsruhe was successfully performed in 2001 and 2002 and has confirmed that the used design and construction principles are applicable for the ITER TF coils.The TFMC was tested up to the rated current of 80 kA as a single coil and in the background field of the EURATOM LCT coil in order to achieve ITER TF coil relevant stress levels. For the operation of the TFMC and LCT coils, special developed forced‐flow‐cooled current leads were used. Both coils with a total weight of 108 t were forced‐flow‐cooled with supercritical He at 4.5 K in a secondary cooling loop connected to the 2 kW refrigerator. However, for currents above 11.4 kA in the LCT coil, its winding had to be cooled at 3.0 K with a separate refrigerator and cooling system.Details of the process engineering of both cooling systems will be described. The o...

The International Thermonuclear Experimental Reactor (ITER) Toroidal Field Conductor Insert (TFCI) has been tested at JAERI Naka, Japan, in 2001, in the background field of the Central Solenoid Model Coil. The TFCI, a well-instrumented /spl sim/43 m long Nb/sub 3/Sn solenoid with a thin Ti jacket, wound inside a SS mandrel and cooled by supercritical helium (SHe) at 4.5 K and 0.6 MPa, was successfully operated up to 46 kA and 13 T. Among others, tests of quench propagation, with delay time of the current dump up to 7 s, were performed driven by an inductive heater. The experimental results of these tests are presented. The hot spot temperature reached in the TFCI during the quench is qualitatively assessed. A more quantitative quench analysis is then performed using the Mithrandir code, confirming the qualitative estimation of the hot spot temperature and showing the importance of heat loss to the mandrel in the slowing down of quench propagation. The computed results reproduce well the main experimental features of the quench transient up to the current dump.

The full set of TCS measurements performed during 2000–2002 on conductor 1A of the ITER Central Solenoid Model Coil (CSMC) of the International Thermonuclear Experimental Reactor (ITER) is analysed with the extensively validated M&M code. Under the assumptions of uniform strand properties and uniform current distribution among strands, the performance of this ‘‘average’’ strand in the cable is deduced from the best fit of the measured voltage–inlet temperature characteristics. Two fitting parameters are chosen: an ad-hoc additional contribution eextra to the longitudinal strain of the average strand in the cable, and the average-strand (or cable ‘‘effective’’) index n of the electric field–current density characteristic. It is shown that the average strand in the CSMC performed less well than expected from the strand database––an increasingly negative eextra being needed to reproduce the coil behaviour at increasing transport currents, and that the cable effective n is clearly below the measured n of the strand. It is argued that this average-strand performance reduction in the CSMC is most likely related to mechanical load effects. � 2003 Elsevier Science Ltd. All rights reserved.

A new 3D CFD model has been developed and applied to the hydraulic analysis of three different inlet options for the ITER Central Solenoid. From the point of view of flow repartition among the petals, it appears that shorter, more compact options have even a marginal advantage over a long slit inlet; their hydraulic impedance is larger, but always within the allowable. No overheating associated to low n value of the Nb3Sn conductor is predicted for either inlet option.

The PFCI is a single-layer solenoid wound from a 45 m long ITER-type NbTi dual-channel cable-in-conduit conductor, designed to be representative of the one currently proposed for the ITER PF1&6 coils. The PFCI, installed in the bore of the ITER central solenoid model coil (CSMC) at JAEA Naka, Japan, and well instrumented from both the thermal hydraulic and the electromagnetic points of view, has been successfully tested in June-August 2008. The test concentrated on DC performance (current sharing temperature and critical current measurements) and AC loss measurements. The results of the analysis of those measurements are reported in the paper, with particular attention to the comparison with the PFCI short sample, which was previously tested in the SULTAN facility. The evolution of the DC performance of the CSMC is also discussed.

In the frame of the International Thermonuclear Experimental Reactor (ITER), several short full-size Nb3Sn samples of candidate toroidal field (TF) conductors were tested in 2007 at the SULTAN facility, PSI Villigen, Switzerland, in conditions relevant to the ITER TF (background magnetic field of 10.78?T and transport current of 68?kA). The performance of a SULTAN sample is determined by the current sharing temperature TCS. This can be obtained in principle from voltage measurements along the conductor sample, but the procedure is not free of issues and ambiguities. Here a complementary approach, based on the calorimetric assessment of the Joule heating due to current sharing, is critically discussed. Suitable algorithms are defined and the respective error bars are estimated, also based on numerical thermal?hydraulic modeling. The calorimetric approach is then applied to assess the performance of the samples tested in 2007 and compared with the results of the standard (electrical) approach.

In 2005-2006, the European Union tested in the SULTAN facility (CRPP Villigen, Switzerland) four ITER-type conductors assembled in two samples. Each conductor made use of an advanced Nb3Sn strand specially developed by European companies for the ITER Toroidal Field (TF) coil conductor. The cable layout of these conductors was the one of the ITER TF Model Coil conductor. In 2007, the European Union tested in SULTAN four other conductors within two other samples as prototypes of the updated ITER TF conductor. These conductors differed mainly from the former by their cable layout whereas identical or similar Nb3Sn strands were used. The paper reports on the current sharing temperature measurements of the second samples under ITER relevant operating conditions. The effect of cycling (1000 cycles) in current under full magnetic field was also investigated. The analysis of the test results is performed with respect to the strand properties using the usual smeared models of the cable. These test results were found to be significantly better than those obtained on the first conductors. The paper discusses the possible causes as well as the consequences of such a performance scattering.

The effect of the choice of Reynolds-Averaged Navier-Stokes (RANS) type turbulence closure on the Computational Fluid Dynamics (CFD) prediction of convective heat losses from the Solar Two central receiver is considered in this paper for a simplified receiver geometry approximated by flat panels. Computed convective losses at steady state are ∼ 2-3% (1%) of the total power absorbed by the receiver, at high (low) wind speed, depending on the turbulence model chosen. The simulation results are consistent with those of available correlations for rough cylinders, if the macroscopic roughness due to the panel edges is accounted for, as well as with the low speed experimental results, within the respective error bars.

In this paper, we discuss the predictive analysis performed in support of the test program of the International Thermonuclear Experimental Reactor (ITER) poloidal field conductor insert (PFCI). A subset of the test program items was considered, with particular emphasis on DC performance and AC losses. The results and implications of the comparison of selected predictions from different laboratories will be presented.

The ITER toroidal field model coil (TFMC) was designed, constructed and tested by the European Home Team in the framework of the ITER research and development program of the Engineering Design Activities (EDA). The project was performed under the leadership of European Fusion Development Activity/Close Support Unit (EFDA/CSU), Ciarching. in collaboration with the European superconductor laboratories and the European industry. The TFMC wits developed and constructed in collaboration with the European industry consortium (AGAN) and Europa Metalli LMI supplied the conductor, The TFMC was tested in the test phase I as single coil and in phase 11 in the background field of the EURATOM LCT coil in the TOSKA facility of the Forschungszentrum Karlsruhe. In phase 1, the TFMC achieved an ITER TF coil relevant current of about 80kA and further representative test results before the end of the EDA. In the more complex test phase [I. the coil was exposed to ITER TF coil relevant mechanical stresses in the winding pack and case. The tests confirmed that engineering design principles and manufacturing procedures are sound and suitable for the ITER TF full size coils. The electromagnetic. thermo hydraulic and mechanical operation parameters agree well with predictions. The achieved Lorentz force on the conductor was about 800 kN/m. That has been equivalent to the Lorentz forces in ITER TF coils. (c) 2005 Elsevier B.V. All fights reserved.

The status of the development of a coupled neutronic/thermal-hydraulic model for the stability and safety analysis of advanced lead-cooled fast fission reactors is presented.

Since year 2009, the joint of the toroidal field (TF) conductor samples tested in the SULTAN facility at PSI Villigen, CH, is solder-filled, so that the helium coolant can only flow axially inside the central channel. The latter is however plugged starting about 40–45 mm downstream of the joint. The helium has then to pass from the central channel into the annular cable region over such a short length that the desired homogeneity of the flow distribution among the petals in the high field region is not guaranteed a priori, since central helix and petal wrappings act as azimuthally non-uniform obstacles to the radial flow. In the paper we first present a geometrical model for estimating this interference to the radial flow, and combine it with a CFD (ANSYS-FLUENT) model of the hydraulic effect of the plug, in order to estimate the mass flow rate imbalance among the petals at the beginning of the plug. This is then used as boundary condition by the THELMA code, to parametrically assess the results of a TCS measurement.

The paper presents the results of a numerical analysis campaign which studies the steady-state behavior and the typical tests (current-sharing temperature and critical current measurements) foreseen for the SULTAN samples of the ITER TF reference conductor, with special emphasis to the current and electric field distribution among and along the sub-cables. In this analysis, the sample geometrical parameters (twist pitch, joint/termination length) and some electrical parameters (joint, termination and inter-bundle resistance) are supposed to range in their design or their measured boundaries, in order to understand their individual effect on the conductor performances and the test conditions.