Development and validation of the 4C thermal–hydraulic model of the ITER Central Solenoid modules

The ITER Central Solenoid (CS) consists of a stack of six modules, each made of 40 pancakes wound with Nb3Sn Cable-In-Conduit Conductors (CICCs) cooled with supercritical helium (SHe). All six modules (plus one spare) are to be individually cold-tested at the General Atomics final test facility in San Diego (USA), in order to check their performance; the first CS Module (CSM1) was tested in early 2020.A test campaign on a CSM Mock-up (CSM MU) wound with 16 dummy pancakes, i.e., with nonsuperconducting (copper) strands, was already carried out in San Diego at the end of 2017, for the commissioning of the test facility. The analysis of the CSM MU experimental data is presented here.Each CSM is a full magnet with 554 turns; it did not have any thermal-hydraulic (TH) or electrical sensors inside the winding due to insulation reasons, so that, e.g., SHe pressure, temperature and mass flow rate, as well as the voltage, were only measured at the ends of selected pancakes.Therefore, it was essential to employ a thermal-hydraulic (TH) model in order to obtain information on the quantities of interest inside the coil, e.g. which was the voltage across the coil at the moment when the current sharing temperature (TCS) was reached for the first time somewhere in that double-pancake (DP) during a TCS test.The TH model of the CSM, developed and implemented in the validated 4C code, and eventually adopted for the test preparation and interpretation, includes some free parameters, i.e., the inter-pancake and inter-turn thermal coupling, whose uncertainty is mainly due to the complex, multi-layer structure of the turn and pancake insulation. The calibration of these parameters is required to correctly capture the TH behavior of the CSM. For this purpose, the results of the experimental campaign on the CSM MU have been used. The detailed topology of the CSM MU is described and implemented here in a dedicated 4C model. Both slow and fast transients are used for the calibration, e.g., quasi-steady state heating of the SHe, entering a single DP and heat slug tests, respectively. It is shown that the transverse heat transfer within the winding pack could be largely overestimated if the ideal heat conduction across a bulk insulation layer is considered. The calibrated model is then validated on the CSM1 test results

identification of the postulated initiating events of accidents occurring in a toroidal field magnet of the eu demo

AbstractThe design of the European Union (EU) DEMO reactor magnet system, currently ongoing within the EUROfusion consortium, will take advantage of the know-how developed during the design and manufacturing of ITER magnets; however, DEMO will suffer some new, more severe challenges, e.g., larger tritium inventory and higher neutron fluence, both having an impact on safety functions accomplished, among the other systems, also by the magnets. For these reasons, and in view of the need to demonstrate a high availability of the reactor (aimed at electricity production), a new, more systematic assessment of the system safety is required. As a contribution in this direction, the initiating events (IEs) of the most critical accident sequences in the EU DEMO magnet system (with special reference to the toroidal field magnets) are identified here, adopting first a functional analysis and then a failure mode, effects, and criticality analysis. In particular, the following are provided: (1) the EU DEMO magnet syste..

Multiscale hydraulic modeling of the ITER TF he inlets during nominal and off-normal operation

In ITER, the supercritical helium (SHe) coolant enters the superconducting toroidal field (TF) coils from the bore of the magnet, with each inlet feeding two adjacent pancakes. Here, as a complement to and extension of experimental measurements performed by other authors, we address the issue numerically through a 3-D computational fluid dynamic ("micro-scale") study of an ITER TF inlet, in both nominal and backflow conditions (e.g., in the case of a quench of the coil). The localized pressure drop at the inlet turns out to be relatively small. Nevertheless, for demonstration purposes of the multiscale approach, suitable correlations for the localized pressure drop are derived and then implemented in a lumped parameter component, to be used in the 4C system code for the "macroscale" analysis of the entire TF coil and related cryogenic cooling loops

Development of the H4C Model of Quench Propagation in the ENEA HTS Cable-In-Conduit Conductor

Experiments on quench propagation in high-current, high field High Temperature Superconducting (HTS) Cable-In-Conduit Conductors (CICCs) for fusion applications are forthcoming. Among the conductor designs to be tested, there is the ENEA slotted core proposal. In order to support the design of the samples and plan the diagnostics to be employed, a 1D multi-region thermal-hydraulic and electric model of the samples has been developed using the H4C code. After an experimental electric characterization, the model is applied to the simulation of quench propagation in the samples. The simulations show how current redistributes among the tapes and the slots. Additionally, they show that the quench protection strategy is suitable to prevent too high hot-spot temperatures

Advanced methods for loss-of-flow accident precursors identification in a superconducting magnet cryogenic cooling circuit

In nuclear fusion systems, such as ITER, Superconducting Magnets (SMs) will be employed to magnetically confine the plasma. A Superconducting Magnet Cryogenic Cooling Circuit (SMCCC) must keep the SMs at cryogenic temperature to preserve their superconductive properties. Thus, a Loss-Of-Flow Accident (LOFA) in the SMCCC is to be avoided. In this work, a three-step methodology for the prompt identification of LOFA precursors (i.e., those component failures leading to a LOFA) is developed. First, accident scenarios are randomly generated by Monte Carlo sampling of the SMCCC components failures and the corresponding transient system response is simulated by a deterministic thermal-hydraulic code. In this phase, fast-running Proper Orthogonal Decomposition (POD)based Kriging metamodels, adaptively trained to mimic the behavior of the detailed long-running code, are employed to reduce the associated computational burden. Second, the scenarios generated are grouped by a Spectral Clustering (SC) embedding the Fuzzy C-Means (FCM), in order to characterize the principal patterns of system evolution towards abnormal conditions (e.g., a LOFA). Third, an On-line Supervised Spectral Clustering (OSSC) approach is developed to assign signals measured during plant operation to one of the prototypical clusters identified, which may reveal the corresponding LOFA precursors (in terms of combinations of failed SMCCC components). The devised method is applied to the simplified model of a cryogenic cooling circuit of a single module of the ITER Central Solenoid. Results show that the approach developed timely identifies 95% of LOFA events and approximately 80% of the corresponding precursors

Mechanical analysis of the ENEA TF coil proposal for the EU DEMO fusion reactor

The design of the superconducting magnet system of the European DEMO fusion reactor is currently being pursued in the framework of the EUROfusion Magnets Work Package (WPMAG). Three alternative winding pack (WP) options for the Toroidal Field Coils (TFCs) are being proposed by different research units, each featuring a different conductor manufacturing technology (react-and-wind vs. wind-and-react) or winding layout (layer vs. pancake). One of the options (namely, WP#2), proposed by Italian ENEA, features a layer-wound WP design adopting a wind-and-react conductor with rectangular cross section with high aspect ratio, obtained squeezing an initially circular conductor. In order to assess the capability of all the TFC components to withstand the electromagnetic loads due to the huge Lorentz forces without any structural failure during the magnet lifetime, the mechanical analysis of the 2016 version of the WP#2 design option is performed here applying a hierarchical approach herein defined as the Stress Recovery Tool (SRT): the Finite Element Analysis (FEA) of a whole magnet (including the casing) is performed at a low computational cost adopting a coarse WP model with smeared (homogenized) properties. The displacements computed on the smeared WP are then used as boundary conditions for a refined FEA of some WP slices, located in selected (critical) poloidal positions, where all the conductors detailed features (jacket, insulations) are properly accounted for

AC Losses in the First ITER CS Module Tests: Experimental Results and Comparison to Analytical Models

The ITER Central Solenoid (CS) will be manufactured by assembling a stack of six modules, which are under fabrication by the US ITER organization and its subcontractors. The tests of the first CS Module have been performed at the premises of the General Atomics (GA) facility in Poway (US), in order to check compliance to the ITER requirements. Among other tests, the magnet was submitted to exponential dumps of the transport current from different initial values (10, 15, 20, 22.5, 25, 35, 40 kA) down to 0 kA. These tests are aimed at conducting DC breaker commissioning of the test facility and were used to measure the AC losses in the coil during electrodynamic transients. This paper presents the results of these measurements, along with a comparison with analytical computations of the losses in the magnet

Cutting-edge R&D activities of CIRTEN in support of the Technology Park annexed to the Italian National Repository of radioactive waste

R&D activities taking place at the institutions belonging to Consorzio Interuniversitario per la Ricerca TEcnologica Nucleare are here presented and discussed. A special focus is on Technology Park annexed to the Italian National Repository of radioactive waste

Advance in the conceptual design of the European DEMO magnet system

The European DEMO, i.e. the demonstration fusion power plant designed in the framework of the Roadmap to Fusion Electricity by the EUROfusion Consortium, is approaching the end of the pre-conceptual design phase, to be accomplished with a Gate Review in 2020, in which all DEMO subsystems will be reviewed by panels of independent experts. The latest 2018 DEMO baseline has major and minor radius of 9.1 m and 2.9 m, plasma current 17.9 MA, toroidal field on the plasma axis 5.2 T, and the peak field in the toroidal-field (TF) conductor 12.0 T. The 900 ton heavy TF coil is prepared in four lowerature-superconductor (LTS) variants, some of them differing slightly, other significantly, from the ITER TF coil design. Two variants of the CS coils are investigated - a purely LTS one resembling the ITER CS, and a hybrid coil, in which the innermost layers made of HTS allow the designers either to increase the magnetic flux, and thus the duration of the fusion pulse, or to reduce the outer radius of the CS coil. An issue presently investigated by mechanical analyzes is the fatigue load. Two variants of the poloidal field coils are being investigated. The magnet and conductor design studies are accompanied by the experimental tests on both LTS and HTS prototype samples, covering a broad range of DC and AC tests. Testing of quench behavior of the 15 kA HTS cables, with size and layout relevant for the fusion magnets and cooled by forced flow helium, is in preparation.</p

The DEMO magnet system – Status and future challenges

We present the pre-concept design of the European DEMO Magnet System, which has successfully passed the DEMO plant-level gate review in 2020. The main design input parameters originate from the so-called DEMO 2018 baseline, which was produced using the PROCESS systems code. It defines a major and minor radius of 9.1 m and 2.9 m, respectively, an on-axis magnetic field of 5.3 T resulting in a peak field on the toroidal field (TF) conductor of 12.0 T. Four variants, all based on low-temperature superconductors (LTS), have been designed for the 16 TF coils. Two of these concepts were selected to be further pursued during the Concept Design Phase (CDP): the first having many similarities to the ITER TF coil concept and the second being the most innovative one, based on react-and-wind (RW) Nb3Sn technology and winding the coils in layers. Two variants for the five Central Solenoid (CS) modules have been investigated: an LTS-only concept resembling to the ITER CS and a hybrid configuration, in which the innermost layers are made of high-temperature superconductors (HTS), which allows either to increase the magnetic flux or to reduce the outer radius of the CS coil. Issues related to fatigue lifetime which emerged in mechanical analyses will be addressed further in the CDP. Both variants proposed for the six poloidal field coils present a lower level of risk for future development. All magnet and conductor design studies included thermal-hydraulic and mechanical analyses, and were accompanied by experimental tests on both LTS and HTS prototype samples (i.e. DC and AC measurements, stability tests, quench evolution etc.). In addition, magnet structures and auxiliary systems, e.g. cryogenics and feeders, were designed at pre-concept level. Important lessons learnt during this first phase of the project were fed into the planning of the CDP. Key aspects to be addressed concern the demonstration and validation of critical technologies (e.g. industrial manufacturing of RW Nb3Sn and HTS long conductors, insulation of penetrations and joints), as well as the detailed design of the overall Magnet System and mechanical structures