Multi-scale thermal-hydraulic modelling for the Primary Heat Transfer System of a tokamak

The EU DEMO reactor is currently in its pre-conceptual design phase by the EUROfusion Consortium members; it aims to be the first tokamak fusion reactor to demonstrate the capability to produce net electrical energy from fusion reactions. To this aim, it must prove tritium self-sufficiency, and so it will be the first tokamak to include a Breeding Blanket (BB), to breed tritium exploiting lithium and the neutrons coming from the fusion reactions. Moreover, to prove feasibility of fusion electricity, the EU DEMO reactor will also be the first to include the power conversion chain, converting the heat coming from fusion reactions into electrical energy, through a Primary Heat Transfer System, which removes the heat deposited in the components close to the plasma and delivering it to the Power Conversion System, that, in the end, produces electricity. Within this framework, a new computational tool is developed, supported by the EUROfusion Programme Management Unit. This code, called the GEneral Tokamak THErmal-hydraulic Model (GETTHEM), aims at fast, system-level, transient thermal-hydraulic modelling of the EU DEMO Primary Heat Transfer System and Balance-of-Plant (BoP), including all the in-vessel and ex-vessel cooling components, and it is the first system-level code of this type explicitly developed for fusion applications. The thermal-hydraulic models of the in-vessel components are developed, starting from the BB, as it is the most thermally loaded component and, consequently, the most important for the BoP. The GETTHEM development currently focuses on two out of the four BB concepts studied in the EU, namely the Helium-Cooled Pebble Bed (HCPB) and the Water-Cooled Lithium-Lead (WCLL) BB concepts. Considering that the EU DEMO is still in pre-conceptual design, the code focuses on execution speed, while maintaining an acceptable accuracy, typically modelling the different components as 0D/1D interconnected objects. GETTHEM is applied to analyse the coolant distribution in the HCPB BB, as well as the maximum temperature reached under normal operating condition in the structural material of both BB concepts, which must stay below 550 °C as a safety requirement. The model is capable to highlight if and where the coolant distribution in the HCPB BB should be optimized in order to avoid an overheating of the structures, allowing at the same time to reduce the compression power needed to circulate the coolant. It also can show if in some regions of the BB, for both coolant options, more detailed analyses are needed, as the current design, tailored on the equatorial BB region, somehow penalizes the regions far from the equatorial plane. Moreover, a separate module of the code is developed, aiming, through suitable simplifications, at fast modelling of accidental transients such as in-vessel Loss-Of-Coolant Accidents (LOCAs). Such module of the code is applied to the parametric analysis of an in-vessel LOCA for HCPB and WCLL, exploiting the code speed to rapidly screen the effect, for instance, of different break sizes, contributing to the proper sizing of the Vacuum Vessel Pressure Suppression System

Thermal-Hydraulic Analysis of the EU DEMO Helium-Cooled Pebble Bed Breeding Blanket Using the GETTHEM Code

The general tokamak thermal-hydraulic model (GETTHEM) has been updated to the most recent version of the EU DEMO helium-cooled pebble bed breeding blanket (BB) design. The GETTHEM results are first benchmarked in a controlled case against the results of 3-D computational fluid-dynamics computations, showing an acceptable accuracy despite the inherent simplifications in the GETTHEM model. GETTHEM is then applied to the evaluation of the poloidal hot spot temperature distribution in an entire BB segment, showing that the maximum temperature in the EUROFER structures overcomes the design limit of 550 °C by more than 50 °C in some blanket modules. A possible mitigation strategy is then proposed and analyzed, based on the idea of cooling the first wall in parallel with the breeding zone, showing that this solution would allow having the EUROFER in its working temperature range in the entire segment, although at the expense of a larger pressure drop

The appeal of neo-fascism in times of crisis. The experience of CasaPound Italia

The present works sets up to analyze the relationship between radical right activism and the unfolding of the financial crisis in Europe, investigating the extent to which the current economic circumstances have influenced right-wing movements’ political supply and repertoires of action. Using the case study of the Italian neo-fascist group CasaPound, and based on a mix of historiography and ethnographic methods, the present work systematically analyzes the ways in which the group tackles the economic crisis. We find that the crisis offers a whole new set of opportunities for the radical right to reconnect with its fascist legacy, and to develop and innovate crisis-related policy proposals and practices. The crisis shapes the groups’ self-understanding and its practices of identity building, both in terms of collective rediscovery of the fascist regime’s legislation, and in terms of promotion of the fascist model as a ‘third way’ alternative to market capitalism. Even more importantly, the financial crisis plays the role of the enemy against which the fascist identity is built, and enables neo-fascist movements to selectively reproduce their identity and ideology within its practices of protest, propaganda, and consensus building

CFD analysis of natural convection cooling of the in-vessel components during a shutdown of the EU DEMO fusion reactor

In view of the large neutron fluence expected in a fusion power plant, the maintenance of the in-vessel components (IVC) must be carried out using Remote Handling (RH); however, before the RH robots can intervene, the temperature of the IVCs must be reduced, so a cooldown phase is required after the reactor shutdown before maintenance activities can start. In the EU DEMO two options are being investigated to cool down the Breeding Blanket (BB) structures before maintenance, namely introducing fans to pump air in forced convection in the plasma chamber (after opening the Vacuum Vessel), or letting the air at room temperature cool down the structures by natural convection; if the required downtime is acceptable, the second option is clearly preferred, as it would reduce the cost and complexity of the system. This work analyses the natural convection option via a 3D transient Computational Fluid-Dynamics (CFD) conjugate heat transfer model, to evaluate the required time to cool down the BB

Hybrid 1D + 2D Modelling for the Assessment of the Heat Transfer in the EU DEMO Water-Cooled Lithium-Lead Manifolds

The European demonstration fusion power reactor (EU DEMO) tokamak will be the first European fusion device to produce electricity and to include a breeding blanket (BB). In the framework of the design of the EU DEMO BB, the analysis of the heat transfer between the inlet and outlet manifold of the coolant is needed, to assess the actual cooling capability of the water entering the cooling channels, as well as the actual coolant outlet temperature from the machine. The complex, fully three-dimensional conjugate heat transfer problem is reduced here with a novel approach to a simpler one, decoupling the longitudinal and transverse scales for the heat transport by developing correlations for a conductive heat-transfer problem. While in the longitudinal direction a standard 1D model for the heat transport by fluid advection is adopted, a set of 2D finite elements analyses are run in the transverse direction, in order to lump the 2D heat conduction effects in suitable correlations. Such correlations are implemented in a 1D finite volume model with the 1D GEneral Tokamak THErmal-hydraulic Model (GETTHEM) code (Politecnico di Torino, Torino, Italy); the proposed approach thus reduces the 3D problem to a 1D one, allowing a parametric evaluation of the heat transfer in the entire blanket with a reduced computational cost. The deviation from nominal inlet and outlet temperature values, for the case of the Water-Cooled Lithium-Lead BB concept, is found to be always below 1.4 K and, in some cases, even to be beneficial. Consequently, the heat transfer among the manifolds at different temperatures can be safely (and conservatively) neglected

Automatic 3D foetal face model extraction from ultrasonography through histogram processing

Ultrasound is by far the most adopted method for safe screening and diagnosis in the prenatal phase, thanks to its non-harmful nature with respect to radiation-based imaging techniques. The main drawback of ultrasound imaging is its sensitivity to scattering noise, which makes automatic tissues segmentation a tricky task, limiting the possible range of applications. An algorithm for automatically extracting the facial surface is presented here. The method provides a comprehensive segmentation process and does not require any human intervention or training procedures, leading from the output of the scanner directly to the 3D mesh describing the face. The proposed segmentation technique is based on a two-step statistical process that relies on both volumetric histogram processing and 2D segmentation. The completely unattended nature of such a procedure makes it possible to rapidly populate a large database of 3D point clouds describing healthy and unhealthy faces, enhancing the diagnosis of rare syndromes through statistical analyses

Parametric thermal-hydraulic analysis of the EU DEMO Water-Cooled Lithium-Lead First Wall using the GETTHEM code

The system-level code GETTHEM is applied to the thermal-hydraulic analysis of an entire segment of the Water-Cooled Lithium-Lead (WCLL) First Wall (FW) of the EU DEMO reactor, parametrically varying the heat load of the FW and the coolant mass flow rate. The results show that the WCLL FW design can tolerate variations of the distribution of the heat flux with respect to the design value, without requiring modifications. The top inboard and the bottom outboard regions are identified as most critical from the point of view of the cooling of the FW. Finally, the largest possible extent of the WCLL FW surface where the peak heat load can be safely applied is identified through a parametric analysis, performed on the critical regions, to understand which is the limit of the cooling capacity of the system

A Transient 3-D CFD Model for the Simulation of Forced or Natural Convection of the EU DEMO In-Vessel Components

As the EU DEMO reactor will act as a Component Test Facility for the breeding blanket (BB), it is foreseen that the different BB concepts will be tested throughout the plant’s lifetime. The maintenance of all the in-vessel components (IVCs), as for all D-T fusion machines, must be carried out employing remote handling (RH) technology, as the structural materials will be activated by the neutrons. The maintained segment and possibly other nearby segments cannot be actively cooled and will heat up due to the decay heat. For these reasons, alternative cooling strategies need thus to be investigated to ensure that the BB segment will cool down within the limits required by the RH in a reasonable amount of time. In the present work, two possible cooling options are investigated for the case of the Water-Cooled Lithium-Lead BB concept. One is based on the passive cool-down by natural convection of the BB segments, whereas the second one relies on a forcing flow of cool air on the BB surfaces. A computational fluid dynamics (CFD) approach has been used to study the different options for performing transient analyses through the Star-CCM+ commercial code

Modelling an in-vessel loss of coolant accident in the EU DEMO WCLL breeding blanket with the GETTHEM code

One of the accidents to be analyzed for the operation of the EU DEMO tokamak reactor is the in-vessel Loss-Of-Coolant Accident (LOCA), in which a postulated rupture in the First Wall causes a rapid pressurization of the Vacuum Vessel (VV). To avoid rupture of the VV, a VV Pressure Suppression System (VVPSS) is used, which is aimed at removing the coolant from the VV, preserving its integrity and safely storing the coolant together with the radioactive products contained therein. A system-level tool for the analysis of thermal-hydraulic transients in tokamak fusion reactors, called GEneral Tokamak THErmal-hydraulic Model (GETTHEM), is under development at Politecnico di Torino. This paper presents the GETTHEM module developed for the description of the EU DEMO VVPSS, in the case of a water-cooled Breeding Blanket concept; the code validation against experimental data coming from the Ingress of Coolant Event campaign performed in Japan is shown. The tool is then applied to a parametric analysis relevant for an EU DEMO in-VV LOCA, and the results are presented and discussed