The water-cooled lead lithium breeding blanket (WCLL BB) is one of two BB candidate concepts to be chosen as the driver blanket of the EU-DEMO fusion reactor. Research activities carried out in the past decade, under the umbrella of the EUROfusion consortium, have allowed a quite advanced reactor architecture to be achieved. Moreover, significant efforts have been made in order to develop the WCLL BB pre-conceptual design following a holistic approach, identifying interfaces between components and systems while respecting a system engineering approach. This paper reports a description of the current WCLL BB architecture, focusing on the latest modifications in the BB reference layout aimed at evolving the design from its pre-conceptual version into a robust conceptual layout. In particular, the main rationale behind design choices and the BB’s overall performances are highlighted. The present paper also gives an overview of the integration between the BB and the different in-vessel systems interacting with it. In particular, interfaces with the tritium extraction and removal (TER) system and the primary heat transfer system (PHTS) are described. Attention is also paid to auxiliary systems devoted to heat the plasma, such as electron cyclotron heating (ECH). Indeed, the integration of this system in the BB will strongly impact the segment design since it envisages the introduction of significant cut-outs in the BB layout. A preliminary CAD model of the central outboard blanket (COB) segment housing the ECH cut-out has been set up and is reported in this paper. The chosen modeling strategy, adopted loads and boundary conditions, as well as obtained results, are reported in the paper and critically discussed.

The latest progress in the design of the water-cooled lithium–lead (WCLL) tritium extraction and removal (TER) system for the European DEMO tokamak reactor is presented. The implementation and optimization of the conceptual design of the TER system are performed in order to manage the tritium concentration in the LiPb and ancillary systems, to control the LiPb chemistry, to remove accumulated corrosion and activated products (in particular, the helium generated in the BB), to store the LiPb, to empty the BB segments, to shield the equipment due to LiPb activation, and to accommodate possible overpressure of the LiPb. The LiPb volumes in the inboard (IB) and outboard (OB) modules of the BB are separately managed due to the different pressure drops and required mass flow rates in the different plasma operational phases. Therefore, the tritium extraction is managed by 6 LiPb loops: 4 loops for the OB segments and 2 loops for the IB segments. Each one is a closed loop with forced circulation of the liquid metal through the TER and the other ancillary systems. The design presents the new CAD drawings and the integration of the TEU into the tokamak building, designed on the basis of an experimental characterization carried out for the permeator against vacuum (PAV) and gas–liquid contactor (GLC) technologies, the two most promising technologies for tritium extraction from liquid metal.

A scaled-down test section representative of an Outboard Segment manifold of the Water-Cooled Lithium Lead Breeding Blanket for the European DEMO has been designed for installation and test in a high- mass flow branch of the W-HYDRA facility, under construction at the premises of ENEA Brasimone Research Center. The test section should confirm the flow repartition recently computed in the different breeding units on the full-scale manifold, validating at the same time the computational tools used for the design and analysis. The detailed objectives and requirements of the test section, as well as the scaling rationale and procedure adopted for its design, are presented in the paper. The final design of the test section is discussed. The preliminary analyses of the developed design are also presented and show that it is compliant with the initial objectives.