Evolution of poloidal variation of impurity density and ambipolar potential in rotating tokamak plasma Part II

We present numerical results of a 1-D (poloidal), time dependent code for the description of ion impurity transport on a given tokamak magnetic surface, in the presence of momentum sources such that, as observed in neutral beam injection experiments, the toroidal rotation velocities be comparable to or larger than the impurity thermal speed. We show that the densities, the velocities and the ambipolar potential reach a quasi steady state characterized by significant poloidal gradients, on a time scale of the order of the collision time, i.e. faster than the radial diffusion scale. To obtain this steady state a phenomenological drag force needs be introduced; we find that a purely classical, gyroviscous force alone is apparently insufficient to obtain a steady state, within the framework of the present model which retains only the zero-th order, in the Larmor radius expansion, ion-impurity friction

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

SOLPS-ITER modeling of ASDEX Upgrade L-mode detachment states

SOLPS-ITER is used to model ASDEX Upgrade L-mode detachment states including the onset of detachment, the fluctuating detachment, and the complete detachment states, considering drifts and mimicking filamentary convective transport with a radial outward velocity in the low field side. The effect of drifts, perpendicular outward convection and core boundary conditions on the numerical solution is presented. The modeling results are validated against experimental data. We find a good agreement of particle flux at the inner target between modeling results and experimental data. On the opposite, at the outer target computations underestimate measured particle flux by a factor of about 2 ∼ 3 in the onset of detachment and the fluctuating detachment states

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

Functional safety assessment of a liquid metal divertor for the European demo tokamak

A reliable strategy for the heat exhaust problem for fusion reactors is among the milestones indicated in EUROfusion (2018). In a fusion reactor, the divertor targets are subject to extremely large heat and particle fluxes. For fusion to be economically feasible, these conditions must be withstood without damage for long time. The “baseline” strategy will be employed for the ITER experiment (which is being built in France) and is based on actively cooled tungsten monoblocks. It is unclear whether this strategy will extrapolate to a future fusion reactor (such as the EU-DEMO, whose pre-conceptual design is ongoing within the EUROfusion consortium). For this reason, alternative solutions are under study, which will eventually be tested in a dedicated experiment in Italy, namely the Divertor Tokamak Test (DTT). One possibility is to employ liquid metal divertors (LMDs), for which the plasma-facing surface is inherently self-healing and immune to thermo-mechanical stresses. Within the framework of the pre-conceptual design of an LMD for the EU-DEMO, safety issues need to be considered at an early stage. In this work we present a preliminary but systematic safety analysis for this system, by means of the Functional Failure Mode and Effect Analysis (FFMEA). The FFMEA allows to identify possible accident initiators for systems undergoing pre-conceptual design, when more specific safety evaluations (e.g. at the component level) are not possible, US Nuclear Regulatory Commission (2009). This is done by postulating the loss of a system function rather than a specific component failure, thus compensating for the lack of detailed design information. For each function, the potential causes of its loss, a plausible evolution and preventive and mitigative measures are investigated, possibly specifying the need for further information. The initiating events are grouped according to consequences and the plant response. For each group, the Postulated Initiating Event (PIE) is chosen. The PIEs list drives and limits the set of accidental scenarios which will undergo deterministic analysis in a successive phase of the work, in order to evaluate the capacity of the system to withstand/mitigate its consequences. This will assess whether safety limits are respected or whether additional safety provisions are required. From the PIEs list, the design basis accident (DBA) and beyond design basis accident (BDBA) will eventually be selected

Analysis of the effects of primary heat transfer system isolation valves in case of in-vessel loss-of-coolant accidents in the EU DEMO

As DEMO is the first European device planned to produce electricity from fusion, the volume of its Primary Heat Transfer Systems (PHTS) will be consistently larger if compared to present or next-generation tokamaks such as ITER. The consequences of an in-vessel Loss-Of-Coolant Accident (LOCA) would then be more important, and within the EUROfusion Consortium different possible mitigation measures are being investigated. Among these, the introduction of Isolation Valves (IsoVs) on the main cooling loops of the Breeding Blanket is being considered, in view of the many benefits they would introduce, not only in case of accidents, but also e.g. during the maintenance of the in-vessel components. Fast-closing IsoVs on the PHTS would help in relaxing not only the requirements of the VV pressure suppression system (VVPSS) design, but also those related to the expansion volumes that shall accommodate the contaminated coolant discharged from the PHTS after a LOCA. In the present work, the GETTHEM code, the system-level thermal-hydraulic model developed for the EU DEMO at Politecnico di Torino, is used to assess the beneficial effects of the introduction of the IsoVs. The effects of the actuation time of the IsoVs and of their location are parametrically investigated, considering both water and helium as PHTS coolants, with particular reference to the reduction of the in-vessel space-averaged pressure and of the suppression system size