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In the International Thermonuclear Experimental Reactor, the nuclear radiation escaping from the vacuum vessel reaches the superconducting toroidal field (TF) magnets, affecting the temperature margin ΔTmar, that is, the difference between the current sharing temperature and the operating temperature. The TF magnets are designed to operate at a minimum margin ΔTmarmin = 0.7 K. Recent design activity on in-vessel components, for example, blanket, in-vessel coils for plasma stability, suggests a potential enhancement of the nuclear heat load, leading to a reduction of ΔTmar, which is accurately assessed in the paper using the validated 4C code. For the case when the margin goes below the minimum, different possible mitigation strategies are investigated: the first considers the possible reduction of the He bath temperature from the nominal 4.3 K down to 3.8 K, and is proven to be successful. The others consider the possible increase of the dwell time between plasma pulses, and is shown to be inadequate, or the decrease of plasma pulse duration, which turns out to be effective below 300 s.

The magnet system of ITER includes high temperature superconducting (HTS) current leads with a maximum current of 68 kA for the toroidal field (TF) coils, 55 kA for the poloidal field (PF)/central solenoid (CS) coils and 10 kA for the control coils (CC), respectively. Although different in terms of size and operative conditions, the ITER HTS current leads have been all designed on the basis of an established concept, which was successfully developed for the LHC at CERN and proven by the so-called 70 kA “demonstrator” lead made by KIT and by the ITER pre-prototypes made by ASIPP in China. A broad R&D campaign has been undertaken by ASIPP and CERN in order to find optimized designs for each component of the leads. Nevertheless, a comprehensive picture of the performance of the entire HTS current leads is not yet available. In this paper, a steady state, full length, thermal-hydraulic 1-D modeling is applied to the study of the three types (TF, PF/CS, CC) of ITER HTS current leads. The results of this predictive analysis are then compared with relevant ITER requirements. It was found that the present design of the HTS current leads will fulfill these specifications.

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 first step of the validation of the recently developed thermal-hydraulic code 4C is presented. We study the transient evolution during a so-called standard 25 kA safety discharge of the International Thermonuclear Experimental Reactor (ITER) Toroidal Field Model Coil (TFMC), tested in 2001-2002 in the TOSKA facility at the Karlsruhe Institute of Technology, Germany. The computed outlet He temperatures from both winding and case cooling channels, as well as the local maximum temperature increase of the structures, are shown to be in good agreement with the measured data.

We analyse the issues of quench propagation in the NbTi Poloidal Field Conductor Insert (PFCI), recently tested at JAEA Naka, Japan. The simulation tools Mithrandir, already validated against data from previous Nb3Sn Insert Coils, and M3, implementing a more detailed thermal-hydraulic description of the CICC cross section, are used. The results of the analysis are reported in the paper and compared with experimental data, with particular attention to NbTi versus Nb3Sn features and to the effects of different model assumptions.

The THELMA code is used to study the coupled thermal-hydraulic electro-magnetic problem of the current and temperature distribution inside the TFPRO2 Nb3Sn SULTAN sample, which was tested in 2007. The code computes self-consistent voltage and temperature values both on the jacket, where they are measured, and inside the cable, where they are more directly representative of the conductor performance. The measured temperature gradients, related to non-uniform Joule heating at the joint and in the high-field region, as well as to non uniform current distribution, are reasonably well reproduced by the model, together with the voltage-current characteristics.

The 1D stability analysis of a subset of the International Thermonuclear Experimental Reactor (ITER) Poloidal Field (PF) coils, namely PF2, PF3, PF5, 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.

The THELMA (Thermal-Hydraulic-ELectro-MAgnetic) code has been developed with the aim of simulating the main aspects of superconductors to be used in the coils of the International Thermonuclear Experimental Reactor (ITER). An application of the code is presented here, where THELMA is used to simulate a single strand by considering, as cable elements, groups of superconducting (SC) filaments and the corresponding portion of the resistive matrix. This approach is used to reproduce the voltage-current characteristic of a Nb3Sn bronze route strand when subject to a bending mechanical load. Attention is focused particularly on the effect of the applied mechanical load on the critical current, which is considered a relevant item in the explanation of the degradation of the coil performance observed in several ITER Model and Insert Coil experiments. The longitudinal strain of the SC filaments is calculated by means of a composite beam model of the strand, taking into account the nonlinear, temperature-dependent material characteristics of the components. The whole load history is simulated, computing first the thermal strain due to cool-down, and then the mechanical strain due to the bending at 4.2 K

The tests of the toroidal field model coil (TFMC) were completed in 2002 in the TOSKA facility of Forschungszentrum Karlsruhe, Germany. Operation reached a combined 80 kA in the TFMC and 16 kA in the LCT coil, resulting in a peak electromechanical load very close to that expected in the full-size ITER TF coils (800 kN/m). Here we concentrate on the measurements of the current sharing temperature T/sub cs/ of the TFMC conductor, possibly the highlight of the whole test campaign. These tests were performed by increasing in steps the helium inlet temperature T/sub in/ in double pancake DP1, resulting in an increasing normal voltage V across the DP1.1 and DP1.2 conductors, and were repeated for several combinations of currents in the TFMC and in the LCT coil. The analysis of the V - T/sub in/ characteristic by means of the M&M code allows to self-consistently deriving an estimate of T/sub cs/, as well as an indirect assessment of the "average" strain state in the conductor. The TFMC isolated strand has also been very recently characterized at different applied uniaxial strain, and preliminary results indicate a stronger reduction of carrying capacity compared to the extrapolation from Summers scaling used in the analysis so far. As a consequence, the performance of the TFMC conductor, as preliminarily re-evaluated here, appears more in line with the strand performance than in previous analysis, although a BI-dependent "degradation" is still present.

The final design for the inlets of the supercritical He (SHe) coolant of the toroidal field (TF) superconducting magnets of the International Thermonuclear Experimental Reactor (ITER) has to be evaluated on the basis of different aspects, including the interest to minimize the associated localized pressure drop. Based on previous experience made on the analysis of the SHe inlets for the ITER superconducting central solenoid, we have developed and applied a computational fluid dynamics (CFD) model to compute the pressure drop versus mass flow rate characteristic in the TF inlet mock-up which was recently tested at CEA Cadarache, France. The cable model is first calibrated and verified against experimental data from short straight samples. The calibrated model is then applied to the inlet geometry and the results of the calculation are compared with the available measurements, showing very good agreement for sufficiently anisotropic permeability of the cable region. The thus validated model is finally used to investigate the distribution of the coolant flow among the different petals and the central channel, downstream of the inlet, as well as to quantitatively assess the role of the petal wrapping on the localized pressure drop at the inlet.

The Karlsruhe Institute of Technology is responsible for the design, construction and testing of the high temperature superconductor (HTS) current leads for the Wendelstein 7-X stellarator and for the JT-60SA tokamak. These HTS current leads mount a heat exchanger of the meander-flow type, in which the helium flows between the fins and is forced to cross flow with respect to the central Cu bar, which actually carries the current. Since an important issue in the operation of the HTS current lead is the optimization of the cooling power consumption, the helium thermal–hydraulics in such complex geometry becomes rather important. In this paper we extend a computational thermal fluid dynamics (CtFD) technique, previously introduced by the same authors and validated on short samples of meander flow heat exchanger and on the W7-X HTS current lead prototype, to a systematic analysis of the helium thermal-fluid dynamics inside different meander flow geometries. The first aim is to clarify under what operative conditions the flow regime can be considered turbulent and how the pressure drop as well as the heat transfer are related to the geometrical parameters and to the flow conditions. From the results of this analysis, correlations for the pressure drop and the heat transfer in the meander flow geometry have also been derived, which are applicable with good accuracy for the design of HXs over a broad range of geometries.

The Karlsruhe Institute of Technology and the Politecnico di Torino have developed and validated a computational thermal-fluid dynamics (CtFD) strategy for the systematic analysis of the thermal-hydraulics inside the meander flow heat exchanger used in high-temperature superconducting current leads for fusion applications. In the recent past, the application of this CtFD technique has shown that some operating conditions occurring in these devices may not reach the turbulent regime region. With that motivation, the CtFD analysis of the helium thermal-fluid dynamics inside different meander flow geometries is extended here to the laminar flow regime. Our first aim is to clarify under which operative conditions the flow regime can be considered laminar and how the pressure drop as well as the heat transfer are related to the geometrical parameters and to the flow conditions. From the results of this analysis, correlations for the pressure drop and for the heat transfer coefficient in the meander flow geometry have been derived, which are applicable with good accuracy to the design of meander flow heat exchangers over a broad range of geometrical parameters.

This paper compares two Computational Fluid Dynamic (CFD) approaches for the analysis of flow boiling inside the first wall (FW) of the International Thermonuclear Experimental Reactor (ITER): (1) the Rohsenow model for nucleate boiling, seamlessly switching to the Volume of Fluid (VOF) approach for film boiling, as available in the commercial CFD code STAR-CCM+, (2) the Bergles–Rohsenow (BR) model, for which we developed a User Defined Function (UDF), implemented in the commercial code FLUENT. The physics of both models is described, and the results with different inlet conditions and heating levels are compared with experimental results obtained at the Efremov Institute, Russia. The performance of both models is compared in terms of accuracy and computational cost.

A new Modelica library has been developed, which provides the framework for the drag-and-drop cryogenic circuit modeling capability of the Cryogenic Circuit Conductor and Coil (4C) code. The code has been applied to the simulation of a thermal-hydraulic transient in a closed supercritical helium (SHe) loop, driven by a pulsed heat load and including multiple feedback controls. The 4C results are in good qualitative agreement with those of the Vincenta code

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.

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.

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.

Two short length samples have been prepared and tested in SULTAN to benchmark the performance of high current density, advanced Nb3Sn strands in the large cable-in-conduit conductors (CICC) for ITER. The cable pattern and jacket layout were identical to the toroidal field model coil conductor (TFMC), tested in 1999. The four conductor sections used strands from OST, EAS, OKSC and OCSI respectively. The Cu:non-Cu ratio was 1 for three of the new strands, compared to 1.5 in the TFMC strand. The conductors with OST and OKSC strands had one Cu wire for two Nb3Sn strands, as in TFMC. In the EAS and OCSI conductors, all the 1080 strands in the cable were Nb3Sn. A dc test under relevant load conditions and a thermal-hydraulic campaign was carried out in SULTAN. The CICC performance was strongly degraded compared to the strand for all the four conductors. The current sharing temperature at the ITER TF operating conditions (jop = 286 A/mm2, B = 11.15 T) was lower than requested by ITER.

The DC performance of the so-called Good Joint Nb3Sn Short Sample conductor, measured in the SULTAN facility in 1999 with a progressively reduced (chopped) joint length, became worse as the joint became shorter. In this paper we present the analysis of those tests using the THELMA code. The results confirm quantitatively that the loss of performance was due to increasingly unbalanced current distribution. Once the unbalance due to the joint is properly modeled, the performance of the conductor can be fitted using the same extra longitudinal strain in the critical current scaling, proportional to the electromagnetic load but independent of the joint length.

The problem of how representative a short Nb3Sn conductor sample is for simulating conductor-in-coil conditions in the ITER magnets is addressed here by considering the particular case of the Central Solenoid Model Coil (CSMC), tested in 2001-2003 at JAERI Naka, Japan, and its associated Good Joint (GJ) sample (tested in 1999 in the SULTAN facility). While the basic thermal-hydraulic conditions are interpreted from experimental data by the M&M code, the sensitivity of the voltage-temperature characteristics to mechanical effects (due to the strain sensitivity of the Nb3Sn) is investigated by analysis. The results are used to assess the errors to be expected in predicting coil performance from measurements made on conductor samples

Bus Bar III (BBIII) is a curvilinear NbTi full-size ITER-type multistrand superconducting cable 7 m long, tested in 2004 at Forschungszentrum Karlsruhe, Germany. The sample was cooled with a forced helium flow and was equipped with many electromagnetic and thermal-hydraulic sensors, to acquire as wide as possible an experimental database, aimed at getting a benchmark for the validation of the THELMA code, a computer tool for the analysis of CICC prototypes and magnets, recently developed and presently under validation. This code can compute the current distribution among the macrostrands used to represent the cable, taking into account their superconducting behavior, as well as the temperature and the magnetic field distribution. The paper deals with the application of THELMA to some tests on BBIII useful for the electromagnetic model validation and presents in detail the analysis setting-up, the results, its critical aspects and limits. A comparison is shown between the measured and the computed waveforms of the electromagnetic data.

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, pick-up 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.

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.

The Bus Bar III (BBIII), fabricated within the Toroidal Field Model Coil Task of the International Thermonuclear Experimental Reactor (ITER), was tested at the Forschungszentrum Karlsruhe, Germany, in the spring of 2004. The BBIII consists of an approximately 7 m long NbTi dual-channel conductor with a thick square stainless steel jacket, cooled by forced flow supercritical He. It was energized with currents up to 80 kA and operates in its self magnetic field (up to /spl sim/0.8 T). The BBIII was instrumented with Hall-probe heads and arrays, voltage rings and longitudinal voltage taps for electro-magnetic measurements, in order to get experimental data to be used for the validation of a recently developed hybrid thermal-hydraulic electro-magnetic code (THELMA), as well as for the assessment of the possibility of performing a reliable reconstruction of the current distribution in the conductor cross section under controlled conditions. In the tests, current ramps at different rates were applied to characterize the conductor time constants, while two different resistive heaters (one upstream of the BBIII inlet, another one directly on the BBIII jacket) were separately operated in order to approach current sharing in the conductor and to observe the related current re-distribution. In this paper, a summary of the collected experimental results is presented, with particular emphasis on those aspects more relevant for the forthcoming THELMA analysis.

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.

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 ~37 K in the winding and ~55 K in the structures. The peak pressurization in the winding remains below ~8.5 MPa.

In this paper we extend the recently developed CtFD strategy for the analysis of the thermal-hydraulics of a meander flow type heat exchanger, validated against the Wendelstein7-X (W7-X) High Temperature Superconductor (HTS) current lead prototype experiment, to the case of a heat exchanger for a 68 kA HTS current lead. We aim to assess the optimized operative conditions and the influence of different geometrical configurations on the performance of the heat exchanger.

As the test of the PFC1 is foreseen in 2006 at JAERI Naka, Japan, it is essential to consider in detail the lessons learned from the short NbTi sample tests, as well as the issues left open after them, in order to develop a suitable test program of the PFC1 aimed at bridging the extrapolation gap between measured strand and future PF coil performance. Here we consider in particular the following issues: 1) the actual possibility to quench the PFCI conductor in the T-CS tests before quenching the intermediate joint, 2) the question of the so-called sudden or premature quench, based on SULTAN sample results, applying a recently developed multi-solid and multi-channel extension of the Mithrandir code to a short sample analysis; 3) the feasibility of the AC losses calorimetry in the PFCI.