Modeling of thermalhydraulic effects of AC losses in the Central Solenoid Insert Coil using the M&M code

Abstract-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 thermalhydraulic effects of AC losses in the coil

Analysis of Quench Propagation in the ITER Poloidal Field Conductor Insert (PFCI)

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

Performance analysis of a graded winding pack design for the EU DEMO TF coil in normal and off-normal conditions

The superconducting magnet system plays an important role in the framework of the design of the EU DEMO tokamak. In recent years, ENEA developed a prototype of cable-in-conduit conductor (CICC) with two low-impedance central channels to be used in the DEMO Toroidal Field (TF) coils with a graded winding pack (WP). In this paper, a model of a TF coil based on the thermal-hydraulic code 4C has been developed, including the WP, the steel casing with dedicated cooling channels (CCCs) and the two independent cryogenic circuits cooling the WP and the casing, respectively. The first part of the work analyzes the performance of the WP during a series of standard plasma pulses in normal operating conditions. In the second part different off-normal operating conditions during the plasma pulses are studied, namely the collapse of one or both central channel(s) in the most critical CICCs and the plugging of some CCCs at the most critical locations in the magnet

Validation of the 4C code against data from the HELIOS loop at CEA Grenoble

We complete the first validation campaign of the Cryogenic Circuit Conductor and Coil (4C) code, focusing on the cryogenic circuit module of 4C. Data from the HELIOS facility (HElium Loop for hIgh LOads Smoothing) at CEA Grenoble, France, are used as reference, together with the component models from the recently developed "Cryogenics" Modelica library. 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 measurement

Dynamic modeling of a Supercritical Helium closed loop with the 4C code

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 cod

Thermal-Hydraulic Simulation of 80 kA Safety Discharge in the ITER Toroidal Field Model Coil (TFMC) using the 4C Code

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

PROGRESS IN MULTI-PHYSICS MODELING OF INNOVATIVE LEAD-COOLED FAST REACTORS

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

4C code analysis of thermal-hydraulic transients in the KSTAR PF1 superconducting coil

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 4 symmetric pairs of coils PF1L/U-…-PF4L/U. Each of the CS coils is pancake wound using Nb3Sn 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 ntau 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 ntau values needed to explain the experimental results decrease at increasing current ramp rates, consistently with the results found in the literature