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In the context of the Work Package DIVertor (WPDIV) of the EUROfusion action, a research campaign has been carried out by University of Palermo in cooperation with ENEA to assess the thermal-hydraulic performances of the DEMO divertor cooling system, concentrating the attention on its 2019 Plasma Facing Components (PFCs) configuration, relevant to DEMO baseline 2017. The research activity has been performed following a theoretical-numerical technique based on the finite volume method and adopting the well-known ANSYS CFX CFD code. The PFCs cooling circuit thermal-hydraulic performances under nominal steady-state conditions, assessed mainly in terms of coolant total pressure drop, coolant axial flow speed and margin against Critical Heat Flux (CHF) distributions among the plasma-facing channels, have been evaluated with a CFD analysis to check their compliance with the corresponding limits. Results have highlighted serious critical issues, such as an intolerable total pressure drop (significantly higher than 1.4 MPa) as well as an insufficient margin against CHF onset (lower than 1.4) within all the PFU channels. Therefore, an optimization study has been performed to investigate the potential improvements of the PFCs cooling circuit thermal-hydraulic performances due to proper changes of its geometric configuration, focussing the attention on the inlet manifold branch. The study has allowed selecting the most effective cooling circuit configuration, that fulfils the maximum pressure drop requirement ( Δ p 1.4 MPa) while raising the minimum CHF margin within the PFU channels. Moreover, it would allow reducing the ex-vessel total pressure drop, while decreasing the average coolant flow velocity up to values lower than 10 m/s. Models, loads and boundary conditions assumed for the analyses are herewith reported and critically discussed, together with the main results obtained.

In the framework of the Work Package DIV 1 – “Divertor Cassette Design and Integration” of the EUROfusion action, a research campaign has been jointly carried out by University of Palermo and ENEA to investigate the steady-state thermal-hydraulic performances of the DEMO divertor cassette cooling system. The research activity has been focussed onto the most recent design of the Cassette Body (CB) cooling circuit, consistent with the DEMO baseline 2017 and equipped with a liner and two Reflector Plates (RPs), whose main functions are to protect the underlying vacuum pump hole from the radiation arising from plasma and shield the PFCs inlet distributors, respectively. The research campaign has been carried out following a theoretical-computational approach based on the finite volume method and adopting the commercial Computational Fluid-Dynamic (CFD) code ANSYS-CFX. The CB thermal-hydraulic performances have been assessed in terms of coolant and structure temperature, coolant overall total pressure drop and flow velocity distribution, mainly in order to check coolant aptitude to provide a uniform and effective cooling to CB, liner and RPs structures. Moreover, the margin against coolant saturation has been evaluated in order check whether any risk of its bulk vaporisation is prevented. The outcomes of the study have shown some criticalities, mainly in terms of coolant bulk vaporisation occurrence at the corner of the Inner Vertical Target (IVT) and uneven coolant flow distribution among RPs plasma-facing channels, that have suggested some design variations whose effectiveness has been numerically assessed. In particular, the solution proposed to contain and reduce the critical region of the IVT corner has been predicted to be particularly effective, while further studies are needed to improve coolant flow distribution among the RPs plasma-facing channels. Models, loads and boundary conditions assumed for the analyses are herewith reported and critically discussed, together with the main results obtained.

The Coolant Purification Systems (CPSs) is one of the most relevant ancillary systems of European Helium Cooled Lead Lithium (HCLL) and Helium Cooled Pebble Bed (HCPB) Test Blanket Modules (TBMs) which are currently in the preliminary design phase in view of their installation and operation in ITER. The CPS implements mainly two functions: the extraction and concentration of the tritium permeated from the TBM modules into the primary cooling circuit and the chemistry control of helium primary coolant. During the HCLL and HCPB-TBSs (Test Blanket Systems) Conceptual Design Review (CDR) in 2015 it was recognized the need of reducing the tritium permeation into the Port Cell #16 of ITER. To achieve this and, then, to lower the tritium partial pressure in the Helium Cooling Systems in normal operation, the helium flow-rate treated by each CPS has been increased of almost one order of magnitude. In 2017, to satisfy the CDR outcomes and the new design requirements requested by Fusion for Energy (F4E, the European Domestic Agency for ITER), ENEA performed a preliminary design of the “enhanced” CPSs. This paper presents the current design of the “enhanced” CPSs, focusing on design requirements, assumptions, selection of technologies and preliminary components sizing.

The validation of the key technologies relevant for a DEMO Breeding Blanket is one of the main objectives of the design and operation of the Test Blanket Systems (TBS) in ITER. In compliance with the main features and technical requirements of the parent breeding blanket concepts, the European TBM Project is developing the HCLL (Helium Cooled Lithium Lead) and HCPB (Helium Cooled Pebble Bed)-TBS, focusing in this phase on the design life cycle and on R&D activities in support of the design. The TBS ancillary systems are mainly circuits devoted to the removal of thermal power and to the extraction and recovery of the tritium generated in the Test Blanket Modules. They are: • The Helium Cooling System (HCS); • The Coolant Purification System (CPS); • The Tritium Extraction System (HCLL-TRS, HCPB-TES); • The Lead Lithium Loop. Their conceptual design was deeply analyzed during the ITER Conceptual Design Review (CDR) in 2015. The assessment of the CDR was overall positive and recommendations for improvements were made. The present design takes into account the main CDR recommendations as well as the implementation of the requirements related to ITER operation, safety principles application and physical space constrains.

Within the framework of the Work Package DIV 1 – “Divertor Cassette Design and Integration” of the EUROfusion action, a research campaign has been jointly carried out by University of Palermo and ENEA to investigate the steady state thermal-hydraulic behaviour of the DEMO divertor cassette cooling circuit, focussing the attention on its Plasma Facing Components (PFCs). The research campaign has been carried out following a theoretical-computational approach based on the Finite Volume Method and adopting the commercial Computational Fluid-Dynamic code ANSYS-CFX. A realistic model of the PFCs cooling circuit has been analysed, specifically embedding each Plasma Facing Unit (PFU) cooling channel with the foreseen swirl tape turbulence promoter, hence resulting in a finite volume model much more detailed than those assessed in previous analyses. Its thermal-hydraulic performances have been numerically evaluated under nominal steady state conditions, also comparing the obtained results with the corresponding outcomes of analogous analyses carried out for a simplified PFCs configuration, without swirl tapes. Moreover, the main thermal-hydraulic parameters have been evaluated in order to check whether the considered PFCs cooling circuit might fulfil the total pressure drop requirement (Δp 1.4). The PFCs cooling circuit thermal-hydraulic behaviour has been additionally assessed at alternative operative conditions, issued to check the viability of a coolant velocity reduction, in order to minimize corrosion and vibrations inside the PFU channels. Models, loads and boundary conditions assumed for the analyses are herewith reported and critically discussed, together with the main results obtained.

Within the framework of the Work Package DIV 1 - “Divertor Cassette Design and Integration” of the EUROfusion action, a research campaign has been jointly carried out by University of Palermo and ENEA to investigate the thermal-hydraulic performances of the DEMO divertor cassette cooling system. The research activity has been focused onto the most recent design of the Cassette Body (CB) cooling circuit equipped with a Liner, whose main function is to protect the underlying vacuum pump hole from the radiation arising from the plasma. The research campaign has been carried out following a theoretical-computational approach based on the Finite Volume Method and adopting the commercial Computational Fluid-Dynamic code ANSYS-CFX. The CB thermal-hydraulic performances have been assessed in terms of coolant and structure temperature, coolant overall pressure drop, flow velocity distribution and mass flow rate fed to the Liner cooling circuit, mainly in order to check coolant aptitude to provide a uniform and effective cooling to both CB and Liner structures. The outcomes of the study have shown some major criticalities, mainly in terms of water coolant vaporization as well as non-symmetric coolant distribution between the two Liner inlets. As a consequence, the following potential solutions have been successfully explored in order to allow the CB to safely operate while complying with its design constraints: • revising the CB design layout in order to increase the coolant mass flow rate fed to Liner; • increasing coolant inlet pressure to rise water saturation temperature and, hence, its margin against vaporization; • increasing coolant mass flow rate to reduce its overall thermal rise. The main results and the achieved optimized model are herewith described and critically discussed.

Often, for the design of a component, several kinds of analyses are needed. Even more frequently, the different fields of study, to be taken into account for the design verification, have to be examined minutely until the final results are satisfying. Furthermore, when geometry modifications are required, for instance to fulfill the component functions, the analyses cycle has to be restarted and an iterative process has to be carried out. This procedure may be time-consuming and herald of errors, in particular if it is demanded to the human activity. Therefore, it is more convenient for the scientific community to adopt a numerical tool that can combine various computational codes. On the base of these considerations, one of the greatest and important challenges for the new design tools is to demonstrate the capability for performing multi-physics analysis in an integrated way. This is a prerequisite, above all, when the component is part of a fusion utility like the Breeding Blanket (BB) in European Demonstration Fusion Power Reactor (DEMO). Indeed, for its design, several fields of analysis are involved such as the neutronics, thermal-hydraulics and the thermo-mechanics. The present work outlines a procedure for their coupling. The main characteristics of this new multi-physics integrated approach are (i) the use of the well-known commercial software, widely employed in the BB design, as well as (ii) the employment of the same geometry definition for all the phenomena studied. An effective application of the aforementioned approach to the pre-conceptual design of the Helium Cooled Pebble Bed (HCPB) and of the Water Cooled Lithium Lead (WCLL) is also provided in this paper. Finally, the achieved results are herewith reported and critically discussed.

Within the framework of EUROfusion R&D activity, a research campaign has been carried out at the University of Palermo, in close cooperation with ENEA labs, in order to assess the thermo-mechanical performances of the DEMO Water-Cooled Lithium Lead (WCLL) inboard blanket equatorial module, whether properly integrated within its whole inboard segment. In particular, a detailed 3D model of this segment, including all the other modules, the back-supporting structure and the attachment system, has been considered in order to realistically simulate the boundary conditions affecting the equatorial module behaviour. The study has been focused on the investigation of the module thermo-mechanical performances under the Over Pressurization (Level D) loading scenario envisaged for the DEMO WCLL breeding blanket as a consequence of a small in-box Loss Of Coolant Accident (LOCA) accident. A theoretical-numerical approach, based on the Finite Element Method (FEM), has been followed and the qualified ABAQUS v. 6.14 commercial FEM code has been adopted. The obtained thermo-mechanical results have been assessed in order to verify their compliance with the design criteria foreseen for the structural material. To this purpose, a stress linearization procedure has been performed along the most critical paths located within the inboard equatorial module structure, in order to check the fulfilment of Level D rules prescribed by the SDC-IC structural design code. The obtained results are herewith presented and critically discussed.

The European DEMO conceptual design foresees four Breeding Blanket (BB) concepts that rely on different cooling and breeding technologies. As DEMO has been conceived to deliver net electricity to the grid, the choice of the blanket coolant plays a pivotal role in the reactor design having a strong influence on plant operation, safety and maintenance. Moreover, the machine pulsed operation makes the BB Primary Heat Transfer System (PHTS) the main hub of the DEMO Balance of Plant (BoP). Within this framework, a study has been carried out to design the Ex-Vessel PHTS of the Helium-Cooled Pebble Bed (HCPB) BB concept. The paper describes criteria and rationale followed with the aim to achieve a simple PHTS design based on the adoption of easy-to-manufacture main components. Results of preliminary thermal-hydraulic calculations carried out to size heat exchangers, piping and compressors are presented and critically discussed. Finally, the evaluation of PHTS key parameters as total pressure drops and total coolant inventory is reported to provide input relevant to the assessment of the impact of helium technology on the design integration and feasibility of DEMO BoP.

Currently, for the EU DEMO, two Breeding Blankets (BBs) have been selected as potential candidates for the integration in the reactor. They are the Water Cooled Lithium Lead and the Helium Cooled Pebble Bed BB concepts. The two BB variants together with the associated ancillary systems drive the design of the overall plant. Therefore, a holistic investigation of integration issues derived by the BB and the installation of its ancillary systems has been performed. The issues related to the water activation due to the 16N and 17N isotopes and the impact on the primary heat transfer systems have been investigated providing guidelines and dedicated solution for the integration of safety devices as isolation valves. The tritium retention and the permeation rates through the blanket and its ancillary systems have been also assessed taking into account different operating points both for the BB and ancillaries and comparing, when possible, the releases with the operating and safety limits. Moreover, the issues related to the tritium start-up inventory as well as the uncertainties on the Tritium Breeding Ratio (TBR) due to the integration of the auxiliary systems within the Vacuum Vessel have been also studied. Finally, the impact of the BB concepts on the safety systems like the Vacuum Vessel Pressure Suppression System is described with a particular focus on the different measures that should be implemented according to the considered concept. All these aspects are then taken into account to drive future developments during the Concept Design Phase., peer-reviewed

In the frame of the activities promoted and encouraged by the EUROfusion Consortium aimed at developing the EU-DEMO fusion reactor, great emphasis has been placed at a very early stage of the design to incorporate the provisions needed to improve the overall plant safety and reliability performances as well as to analyse possible mitigation actions. In this framework, the research activity has been focused on the representative and safety relevant cooling loop of the Helium Cooled Pebble Bed (HCPB) Breeding Blanket (BB) Primary Heat Transfer System (PHTS), purposely selected by the safety team, in order to assess its thermal-hydraulic behaviour during normal operational conditions (ramp up/down and steady state) and to preliminarily investigate the consequences of an ex-vessel LOCA accidental scenario ensuing a Double-Ended Guillotine (DEG) break in the hot leg. The research activity has been carried out following a theoretical-computational approach based on the finite volume method adopting the RELAP5−3D system code along with the ANSYS CFX computational fluid dynamic code, which were implicitly integrated to achieve a more detailed and realistic simulation of the EU-DEMO reactor thermal-hydraulics. Models, assumptions and outcomes of this preliminary study are herein presented and critically discussed.

The Water-Cooled Lead–Lithium Breeding Blanket (WCLL BB) is one of the two blanket concept candidates to become the driver blanket of the EU-DEMO reactor. The design was enacted with a holistic approach. The influence that neutronics, thermal-hydraulics (TH), thermo-mechanics (TM) and magneto-hydro-dynamics (MHD) may have on the design were considered at the same time. This new approach allowed for the design team to create a WCLL BB layout that is able to comply with different foreseen requirements in terms of integration, tritium self-sufficiency, and TH and TM needs. In this paper, the rationale behind the design choices and the main characteristics of the WCLL BB needed for the EU-DEMO are reported and discussed. Finally, the main achievements reached during the pre-conceptual design phase and some remaining open issues to be further investigated in the upcoming conceptual design phase are reported as well.

The Water Cooled Lithium Lead Test Blanket System (WCLL TBS) is one of the EU Test Blanket Systems candidate for being installed and operated in ITER. In view of its Conceptual Design Review by F4E and ITER Organization (IO), planned for mid-September 2020, several technical activities have been performed in the areas of WCLL TBS Ancillary Systems design. In this article the outcomes of the conceptual design phase of the four main Ancillary Systems of WCLL TBS, namely the Water Cooling System (WCS), the Coolant Purification System (CPS), the PbLi loop and the Tritium Extraction System (TES), are reported and critically discussed. In particular, for each Ancillary System hereafter are reported: i) a short design description, including the conceptual design of their main components together with their operative conditions under the so-called Normal Operational State (NOS), ii) the ESP-ESPN classification for their main components, and iii) their arrangement and integration in the assigned ITER areas (PC#16, Vertical Shaft, TCWS Vault, Galleries and Tritium Process Room).

In the frame of the activities promoted and encouraged by the EUROfusion Power Plant Physics and Technology (PPPT) department aimed at developing the EU-DEMO fusion reactor, strong emphasis has been recently addressed to the whole Balance of Plant (BoP) which represents the set of systems devoted to convert the plasma generated thermal power into electricity and to deliver it to the grid. Among these systems, the Divertor Primary Heat Transfer Systems (PHTSs) are intended to feed coolant to the two main components of the Divertor assembly, namely the Plasma Facing Components (PFCs) and the Cassette Body (CB). Since the DEMO Divertor must withstand high heat flux loads together with a considerable neutron deposited power, very tight tolerances may be allowed to the coolant inlet conditions. Therefore, the design of reliable PHTSs is of the utmost importance towards the development of an EU-DEMO fusion reactor. Within this framework, a study has been jointly carried out by University of Palermo, Ansaldo Nucleare and CREATE to design the Ex-Vessel PHTSs of both the PFCs and the CB for a DEMO reactor equipped with a Helium Cooled Pebble Bed Breeding Blanket. The paper describes criteria and rationale followed with the aim to achieve simple PHTS designs based on the adoption of easy-to-manufacture main components avoiding too much extrapolation from the state-of-art technology. Results of preliminary thermal-hydraulic calculations carried out to size heat exchangers, pressurizers, piping and pumps are presented and critically discussed, with particular attention to those integration, safety and feasibility constraints that may deeply affect the design of such components. Finally, the evaluation of PHTS key parameters as total pumping power and coolant inventory is reported.

In the frame of the pre-conceptual design activities of the DEMO work package DIV-1 “Divertor Cassette Design and Integration” of the EUROfusion program, a mock-up of the divertor outer vertical target (OVT) was built, mainly in order to: (i) demonstrate the technical feasibility of manufacturing procedures; (ii) verify the hydraulic design and its capability to ensure a uniform and proper cooling for the plasma facing units (PFUs) with an acceptable pressure drop; and (iii) experimentally validate the computational fluid-dynamic (CFD) model developed by the University of Palermo. In this context, a research campaign was jointly carried out by the University of Palermo and ENEA to experimentally and theoretically assess the hydraulic performances of the OVT mock-up, paying particular attention to the coolant distribution among the PFUs and the total pressure drop across the inlet and outlet sections of the mock-up. The paper presents the results of the steady-state hydraulic experimental test campaign performed at ENEA Brasimone Research Center as well as the relevant numerical analyses performed at the Department of Engineering at the University of Palermo. The test facility, the experimental apparatus, the test matrix and the experimental results, as well as the theoretical model, its assumptions, and the analyses outcomes are herewith reported and critically discussed.

International audience; In a power plant scale fusion reactor, a huge amount of thermal power produced by the fusion reaction and external heating must be exhausted through the narrow area of the divertor targets. The targets must withstand the intense bombardment of the diverted particles where high heat fluxes are generated and erosion takes place on the surface. A considerable amount of volumetric nuclear heating power must also be exhausted. To cope with such an unprecedented power exhaust challenge, a highly efficient cooling capacity is required. Furthermore, the divertor must fulfill other critical functions such as nuclear shielding and channeling (and compression) of exhaust gas for pumping. Assuring the structural integrity of the neutron-irradiated (thus embrittled) components is a crucial prerequisite for a reliable operation over the lifetime. Safety, maintainability, availability, waste and costs are another points of consideration.In late 2020, the Pre-Conceptual Design activities to develop the divertor of the European demonstration fusion reactor were officially concluded. On this occasion, the baseline design and the key technology options were identified and verified by the project team (EUROfusion Work Package Divertor) based on seven years of R&D efforts and endorsed by Gate Review Panel.In this paper, an overview of the load specifications, brief descriptions of the design and the highlights of the technology R&D work are presented together with the further work still needed.

The Water-Cooled Lithium-Lead Breeding Blanket is a key component of a fusion power plant, in charge of ensure Tritium production, shield Vacuum Vessel and magnets and remove the heat power deposited by radiation and particles arising from plasma. The last function is fulfilled by First Wall and Breeding Zone independent cooling systems. Several layouts of BZ coolant system have been investigated in the last years to identify a configuration that might guarantee EUROFER temperature below the limit (550 °C) and good thermal-hydraulic performances (i.e. water outlet temperature of 328 °C). A research activity is conducted to study and compare different modelling approaches to simulate the heat transfer within the BZ liquid metal, assessing their impact on the numerical prediction of the WCLL blanket thermal performances. An approach will rely on the simulation of convective and diffusive heat transfer processes taking place within the liquid metal by means of a CFD tool based on the Finite Volume Method. Conversely, the other approach will roughly assume a pure diffusive heat transfer mechanism within the breeder, due to the very low velocities envisaged for its flow field. In this case the heat transfer performances will be preferably assessed by means of a commercial code based on the Finite Element Method. The analyses have been carried out with reference to the so called “WCLL BB 2018 V0.6″ equatorial cell. Advantages and issues from the thermal-hydraulic point of view are identified, the impact of the imposed boundary conditions and heat transfer properties, with the implemented correlations, on the respective results is critically discussed.

Within the activities promoted by the EUROfusion consortium in support of the design and construction of the DEMO Oriented NEutron Source (DONES), a mock-up of its Target Assembly (TA), based on the configuration with a “bayonet” Back-Plate (BP) and available at ENEA Brasimone labs, is being adopted for the execution of experimental activities aiming at the validation of specific aspects of the target design. Despite the “integral”-TA concept is the current reference, experimental tests concerning the TA pre-heating phase are not significantly affected by the TA concept and are still representative even though conducted on the bayonet-TA concept. Indeed, the main objective of the pre-heating phase is to raise the TA temperature up to a value close to that of the lithium flow in nominal condition (250 °C). In particular, one of the main concerns is to keep the BP at a temperature higher than 200 °C to avoid possible local lithium freezing. In order to support the afore-mentioned tests, a numerical research campaign has been launched by ENEA Brasimone in collaboration with the University of Palermo with the goal of determining an optimal design of the electrical heaters, both in terms of temperature set-points and geometrical layout. To this purpose, several configurations have been assessed following a theoretical-computational approach based on the Finite Element Method (FEM) and adopting the Abaqus FEM code. In a first phase, different electric heater layouts were assessed by means of steady state thermal analyses. Once the most promising scenario was selected, a detailed thermal transient analysis was carried out. An iterative procedure, based on the analysis of the maximum temperature achieved within the mock-up most critical components, was followed in order to determine the duty cycle of the electric heaters. The obtained numerical results are herewith reported and critically discussed.

Within the framework of the activities foreseen by the EUROfusion action on the cooling water activation assessment for a DEMO reactor equipped with a Water Cooled Lithium Lead Breeding Blanket (WCLL BB), the University of Palermo is involved in the assessment of dose rates induced by the decay of nitrogen radioisotopes produced by water activation, nearby the main components (e.g. isolation valves) of both First Wall (FW) and Breeder Zone (BZ) cooling circuits. In particular, the aim of this work is to evaluate the spatial distribution of nitrogen isotopes (16N and 17N) in the WCLL BB cooling circuits. To this purpose, a coupled neutronic/fluid-dynamic problem is solved following a theoretical-numerical approach and adopting an integrated computational tool mainly relying on the use of MCNP6 and ANSYS CFX codes. The operative procedure adopted foresees the assessment of the production rate distributions of nitrogen isotopes within FW and BZ cooling channels and tubes by means of totally heterogeneous neutronic analyses. A fully 3-D fluid-dynamic approach is, then, used to compute the nitrogen isotope concentrations within the In-Vessel complex flow domain, while a 1-D lumped parameters approach is adopted to calculate their distribution along the Ex-Vessel BB Primary Heat Transfer System. The results obtained, herewith presented and critically discussed, provided the necessary data to perform dedicated neutronic and photonic transport analyses and, hence, to assess the dose rates in the aforementioned target locations.

The development of a neutron source able to reproduce the irradiation conditions typical of a nuclear fusion reactor, in order to test candidate structural materials, is the main goal of the Work Package Early Neutron Source (WPENS) of the EUROfusion action. This source, named Demo Oriented NEutron Source (DONES), is a facility where neutrons are produced by means of D-Li interactions. More in detail, a beam of 125 mA deuterium ions at the energy of 40 MeV strikes a lithium jet flowing in a purposely shaped channel in order to obtain an intense and stable neutron flux for the irradiation of material samples. In the framework of these activities, safety analyses are a key aspect in the DONES design and development. Among the postulated initiating events identified during the preliminary Failure Mode Analysis, the Loss of Flow Accident (LOFA) in the Primary Heat Removal System of the lithium loop, due to a trip of the electro-magnetic pump, is one of the most severe. In fact, the loss of lithium flow, combined with the failed stop of the accelerator, could lead to the destruction of the lithium flow channel in correspondence of the component named Back-Plate. For this reason, it has been chosen to investigate the LOFA adopting the deterministic system code RELAP5-3D. Results, obtained from this analysis, are critically discussed and compared with those obtained by a similar calculation carried out with MELCOR 1.8.6 code, in order to assess the RELAP5-3D capability of describing systems adopting lithium as working fluid.

In the frame of the EUROfusion consortium activities the Helium Cooled Pebble Bed (HCPB) and the Water Cooled Lithium Lead (WCLL) concepts are being developed as possible candidates to become driver Breeding Blanket (BB) for the EU DEMO, which aims at the tritium self-sufficiency and net electricity production. The two BB design options encompass water or helium as coolants and solid ceramic with beryllium/beryllides or PbLi as tritium breeder and neutron multipliers. The BB segments have evolved towards a more stable conceptual design taking into account multiple feasibility aspects and requirements imposed by interfacing systems. Possible solutions to improve shielding capabilities of Helium cooled BB are investigated and the impact of water coolant activation is assessed by studying the spatial distribution of 16N and 17N isotopes dose rates, in particular in proximity of isolation valves. The reference and back-up technologies for the Tritium Extraction and Removal (TER) from the helium purge gas and the PbLi are developed addressing key feasibility aspects and implications on the tokamak layout and with considerable R&D efforts. As the BB internals offer an ideal environment (high temperatures, thin structural material) to promote the tritium permeation, studies are devoted to establish a tritium balance in the different systems during operation, with special care to the permeation rate and inventory in the coolant. Those are the key parameters for the feasibility assessment and technology selection for the Coolant Purification Systems (CPS).

In the framework of the Work Package DIV 1 - “Divertor Cassette Design and Integration” of the EUROfusion action, a research campaign has been jointly carried out by University of Palermo and ENEA to investigate the steady-state thermal-hydraulic performances of the DEMO divertor cassette cooling system. The research activity has been focussed onto the most recent design of the Cassette Body (CB) cooling circuit, consistent with the DEMO baseline 2017 and equipped with a liner, whose main function is to protect the underlying vacuum pump CB opening from plasma radiation. The research campaign has been carried out following a theoretical-computational approach based on the finite volume method and adopting the commercial Computational Fluid-Dynamic (CFD) code ANSYS-CFX. The CB thermal-hydraulic performances have been assessed in terms of coolant and structure temperature, coolant overall total pressure drop and flow velocity distribution, mainly in order to check its aptitude to provide a uniform and effective cooling to both CB and liner structures. Moreover, the margin against coolant saturation has been evaluated in order investigate whether any risk of its bulk vaporisation is prevented. The outcomes of the study have shown some criticalities, mainly in terms of structure maximum temperature and coolant vaporization occurrence within the liner. As a consequence, some minor design variations have been suggested within the paper. Models, loads and boundary conditions assumed for the analyses are herewith reported and critically discussed, together with the main results obtained.

Water-cooled lithium-lead breeding blanket is considered a candidate option for European DEMO nuclear fusion reactor. ENEA and the linked third parties have proposed and are developing a multi-module blanket segment concept based on DEMO 2015 specifications. The layout of the module is based on horizontal (i.e. radial-toroidal) water-cooling tubes in the breeding zone, and on lithium lead flowing in radial-poloidal direction. This design choice is driven by the rationale to have a modular design, where a basic geometry is repeated along the poloidal direction. The modules are connected with a back supporting structure, designed to withstand thermal and mechanical loads due to normal operation and selected postulated accidents. Water and lithium lead manifolds are designed and integrated with a consistent primary heat transport system, based on a reliable pressurized water reactor operating experience, and the lithium lead system. Rationale and features of current status of water-cooled lithium-lead breeding blanket design are discussed and supported by thermo-mechanics, thermo-hydraulics and neutronics analyses. Open issues and areas of research and development needs are finally pointed out.

In the framework of EUROfusion Work Package International Cooperation R&D activities, a close collaboration has been started among University of Palermo, ENEA Brasimone and ENEA Frascati for the development of the Water Cooled Lithium Lead (WCLL) Breeding Blanket (BB) concept. In this context, a research campaign has been carried out at the University of Palermo in order to investigate the influence of First Wall (FW) cooling water configuration on the nuclear response of the WCLL BB under irradiation in EU-DEMO, in order to gain useful indications for the WCLL BB pre-conceptual designs. To this end, three-dimensional nuclear analyses have been performed by MCNP5 v. 1.6 Monte Carlo code. A semi-heterogeneous MCNP model of the WCLL BB “single module segment” concept has been used and its nuclear response has been assessed for different FW configurations, centring the attention, mainly, on global quantities as nuclear power deposition and tritium breeding ratio. The results obtained provided some information on the nuclear features of the WCLL BB concept, useful for its design optimisation. The outcomes of this parametric study are herein presented and critically discussed.

In the framework of the EU fusion roadmap implementing activities, an accelerator-based Li(d,xn) neutron source called DONES (Demo-Oriented early NEutron Source) is being designed within the EUROfusion Work Package Early Neutron Source as an essential irradiation facility for testing candidate materials for DEMO reactor and future fusion power plants. DONES will employ a high speed liquid lithium jet struck by a 125 mA, 40 MeV deuteron beam to generate the intense neutron flux used to irradiate the material samples up to the desired level of displacement damage (˜10 dpa/fpy for iron in 0.3 l) and He production rates (˜10-13 appm He/dpa). In order to rapidly achieve a sound and stable design, a new configuration of the DONES target system based on the so-called integral concept has been proposed as reference solution in place of the former baseline design that envisaged a target assembly endowed with a replaceable back-plate, being the latter solution not yet fully qualified and thus not readily implementable. Moreover, following the outcomes of detailed dedicated analyses taking into account different aspects, the design of several components (Quench Tank, Lithium inlet pipe, Vacuum Chamber, etc.) has been changed. In this paper, a brief description of the current design of the DONES Target System is presented including all the above-mentioned aspects, showing the capability of the system to fulfil the prescribed requirements.

The wide range of requirements and constraints involved in the design of nuclear components for fusion reactors makes the development of multi-physics analysis procedures of utmost importance. In the framework of the European DEMO project, the Karlsruhe Institute of Technology (KIT) is dedicating several efforts to the development of a multi-physics analysis tool allowing the characterization of breeding blanket design points which are consistent from the neutronic, thermal-hydraulic and thermal-mechanical points of view. In particular, a procedure developed at KIT is characterized by the implementation of analysis software only. A preliminary step for the validation of such a procedure has been accomplished using a dedicated model of the DEMO Helium Cooled Pebble Bed Blanket 4th outboard module. A global model representative of nuclear irradiation in DEMO and two local models have been set up. Nuclear power deposition and the spatial distribution of its volumetric density have been calculated using Monte Carlo N-Particle transport code for the aforementioned models and compared in order to validate the procedure set up. The outcomes of this comparative study are herein presented and critically discussed.

Within the framework of EUROfusion R&D activities a research campaign has been performed at CEA-Saclay, in close collaboration with the University of Palermo, in order to investigate thermal and thermomechanical performances of the “Optimized Conservative” concept of DEMO Helium-Cooled Lithium Lead breeding blanket (HCLL). Attention has been paid to the HCLL outboard equatorial module (OEM) when subjected to the steady state nominal loading scenario. To this purpose three simplified 3D models, characterized by an increasing level of detail, have been set-up taking into account, firstly, a single radial-toroidal slice, then a basic module geometric unity composed by two adjacent slices and adding, lastly, the peripheral poloidal region. This latter 3D model has allowed the assessment of the Caps potential influence on the module thermal and thermomechanical behaviour. For each investigated 3D model, thermal and thermomechanical analyses have been performed and a stress linearization procedure has been carried out in order to verify the fulfilment of the criteria prescribed by the RCC-MRx 2015 code. The study has been performed adopting a numerical approach, based on the Finite Element Method (FEM), and adopting the Siemens NX v. 10.0 software in order to discretize the geometric domain, whereas thermal and thermomechanical calculations have been carried out using the Cast3 M 2015 FEM code. The obtained results, herewith reported and critically discussed, allow predicting a good thermal and mechanical behaviour of the “Optimized Conservative” concept of DEMO HCLL OEM, even if some small modifications to the module cooling scheme should be performed in order to avoid the insurgence of hotspots where temperature is slightly above the Eurofer limit temperature (550 °C). This will entail, from the mechanical point of view, a reduction of the secondary stress amount which is the main responsible of the failure in RCC-MRx criteria verification within First Wall-Side Wall bend region.

Within the framework of the EUROfusion activities regarding the EU-DEMO Breeding Blanket (BB) concept, the University of Palermo is long-time involved, in close cooperation with ENEA, in the design of the Water Cooled Lithium Lead (WCLL) BB, that is one of the two concepts under consideration for the DEMO reactor. It is mainly characterized by a liquid lithium-lead eutectic alloy acting as breeder and neutron multiplier, as well as by subcooled pressurized water flowing as coolant under PWR-like conditions (pressure of 15.5 MPa and inlet/outlet temperatures of 295 °C/328 °C). A research campaign has been recently carried out to study the potential influence of the Breeding Zone cooling circuit layout, composed of bundles of Double Walled Tubes (DWTs), on the blanket thermo-mechanical behaviour. In the first phase, a parametric study has been performed to investigate the effects of some geometrical DWT parameters onto the blanket thermal response in order to identify the best tubes layout, in case arranged along toroidal-radial planes. In the second phase, the thermo-mechanical performances of the blanket equatorial region, equipped with the optimized DWTs layout, have been assessed. The study has been carried out following a theoretical-numerical approach based on the Finite Element Method and adopting the ABAQUS v. 6.14 commercial code. Model, assumptions and results obtained are herewith presented and critically discussed.

Within the framework of the Work Package DIV 1 - “Divertor Cassette Design and Integration” of the EUROfusion action, a research campaign has been jointly carried out by University of Palermo and ENEA to investigate the thermal-hydraulic behaviour of the DEMO divertor cassette cooling system, focussing the attention on the 2018 configuration of the Plasma Facing Components (PFCs) circuit consistent with the DEMO baseline 2017. The research campaign has been carried out following a theoretical-computational approach based on the finite volume method and adopting the commercial Computational Fluid-Dynamic (CFD) code ANSYS CFX. A steady-state CFD analysis has been carried out for the PFCs cooling circuit under nominal conditions and its thermal-hydraulic performances have been assessed in terms of coolant total pressure drop, flow velocity and Critical Heat Flux (CHF) margin distributions among the Plasma Facing Units (PFUs) channels, to check whether they comply with the corresponding limits. Results obtained have clearly predicted a total pressure drop of ≈1.7 MPa, slightly higher than the prescribed limit of 1.4 MPa, as well as a sufficient margin against CHF onset, higher than 1.4 except for three IVT PFU channels where a minimum value of 1.39 is predicted. Moreover, a parametric study has been performed in order to assess the operating map of the cooling circuit in the phase-space of coolant inlet temperature, pressure and mass flow rate, to be intended as that domain where the circuit thermal-hydraulic performances let it stay within the prescribed requirements. Models, loads and boundary conditions assumed for the analyses are herewith reported and critically discussed, together with the main results obtained.

Within the framework of the activities foreseen by the EUROfusion action on the cooling water activation assessment for a DEMO reactor equipped with a Water Cooled Lithium Lead Breeding Blanket (WCLL BB), the University of Palermo is involved in the investigation of the absorbed dose induced by the decay of nitrogen radioisotopes produced by water activation, in the main components (e.g. isolation valves) of both First Wall (FW) and Breeder Zone (BZ) cooling circuits. The aim of this work is to assess the spatial distribution of the absorbed dose in the DEMO Upper Pipe Chase (UPC), focusing the attention on the space neighbouring a typical isolation valve of the Primary Heat Transfer System (PHTS), taking into account a period of 7 full power years. To this end, a computational approach has been followed adopting MCNP5 Monte Carlo code. In particular, a totally heterogeneous neutronic model of a portion of the UPC has been set up, including the valve and the main FW and BZ PHTS piping, and the spatial distribution of nitrogen isotopes concentrations, previously assessed, have been used to model the photonic and neutronic sources. The results obtained, herewith presented and critically discussed, provided some information on the nuclear issues of the WCLL BB PHTS, to be considered as hints for the blanket design optimization.

In Europe (EU), in the frame of the EUROfusion consortium activities, four Breeding Blanket (BB) concepts are being developed with the aim of fulfilling the performances required by a near-term fusion power demonstration plant (DEMO) in terms of tritium self-sufficiency and electricity production. The four blanket options cover a wide range of technological possibilities, as water and helium are considered as possible coolants and solid ceramic breeder in combination with beryllium and PbLi as tritium breeder and neutron multipliers. The strategy for the BB selection and operation has to account for the challenging schedule of the EU DEMO, the ambitious operational requirements of the BBs and the still large development needed to have a BB qualified and licensed for operating in DEMO. In parallel to the continuous design efforts on the four blanket concepts, their integration in-vessel and ex-vessel has started. On the one hand it has become clear that despite the numerous systems to be integrated in-vessel the protection of the blanket first wall has to be addressed with highest priority. On the other hand the ex-vessel interfaces and the requirements imposed by the blanket to the primary heat transfer system and to the PbLi loop components have a considerable impact on the whole DEMO Plant layout. The aim of this paper is: to present the strategy for the DEMO BB down selection and BB operation in DEMO; to describe the status of the design evolution of the four EU BB concepts; to provide an overview of the challenges of the in-vessel and ex-vessel integration of the main systems interfacing the BBs and describe their design status.

Within the framework of the Work Package Divertor, Subproject: Cassette Design and Integration (WPDIV-Cassette) of the EUROfusion action, a research campaign has been jointly carried out by ENEA and University of Palermo to investigate the thermal-hydraulic performances of the DEMO divertor cassette cooling system. Attention has been focussed on the divertor Plasma Facing Components (PFCs) cooling circuit and a parametric analysis has been carried out in order to assess the potential impact of proper layout changes on its thermal-hydraulic performances, mainly in terms of coolant total pressure drop, flow velocity distribution and margin against critical heat flux occurrence. The research activity has been carried out following a theoretical-computational approach based on the finite volume method and adopting a qualified Computational Fluid-Dynamic (CFD) code. Results obtained have allowed to select a revised PFCs cooling circuit configuration, suitable to comply with the prescribed thermal-hydraulic limits assumed for the DEMO divertor design. They are reported and critically discussed.

Within the framework of the Work Package DIV 1 – “Divertor Cassette Design and Integration” of the EUROfusion action, a research campaign has been jointly carried out by ENEA and University of Palermo to investigate the thermal-hydraulic performances of the DEMO divertor cassette cooling system. A comparative evaluation study has been performed considering the two different options under consideration for the divertor cassette body coolant, namely subcooled pressurized water and helium. The research activity has been carried out following a theoretical-computational approach based on the finite volume method and adopting a qualified Computational Fluid-Dynamic (CFD) code. CFD analyses have been carried out for the considered options of cassette body cooling circuit under nominal steady state conditions and the pertaining thermal-hydraulic performances have been assessed in terms of overall coolant thermal rise, coolant total pressure drop, flow velocity and pumping power, to check whether they comply with the corresponding limits. Results obtained are reported and critically discussed.

Within the framework of the Work Package DIV 1 – “Divertor Cassette Design and Integration” of the EUROfusion action, a research campaign has been jointly carried out by ENEA and University of Palermo to investigate the thermal-hydraulic performances of the DEMO divertor cassette cooling system. A comparative evaluation study has been performed considering three different options for the cooling circuit layout of the divertor Plasma Facing Components (PFCs). The potential improvement in the thermal-hydraulic performance of the cooling system, to be achieved by modifying cooling circuit layout, has been also assessed and discussed in terms of optimization strategy. The research activity has been carried out following a theoretical-computational approach based on the finite volume method and adopting a qualified Computational Fluid-Dynamic (CFD) code. Results obtained are reported and critically discussed.

Within the framework of EUROfusion R&D activities an intense research campaign has been carried out at the University of Palermo, in close cooperation with ENEA Brasimone, in order to investigate the thermo-mechanical performances of the DEMO water-cooled lithium lead breeding blanket (WCLL). In particular, attention has been paid to the most recent geometric configuration of the DEMO WCLL outboard equatorial module, as designed by WCLL project team during 2015, endowed with an attachment system based on the use of radial pins, purposely outlined to connect the module back-plate to its back-supporting structure, that have been properly considered to simulate more realistically the module thermo-mechanical behaviour. The research campaign has been mainly focused on the investigation of the module thermo-mechanical performances under the Normal Operation (Level A) and Over Pressurization (Level D) steady state loading scenarios envisaged for the DEMO WCLL. A theoretical-numerical approach, based on the Finite Element Method (FEM), has been followed and the qualified ABAQUS v. 6.14 commercial FEM code has been adopted. Thermo-mechanical results obtained have been assessed in order to verify their compliance with the design criteria foreseen for the structural material. To this purpose, a stress linearization procedure has been performed along the most critical paths located within the module structure, in order to check the fulfilment of both Level A and Level D rules prescribed by the SDC-IC structural design code. Results obtained are herewith presented and critically discussed, highlighting the open issues and suggesting the pertinent modifications to DEMO WCLL outboard equatorial module design aimed to obtain the complete fulfilment of the prescribed design criteria.

Since 2014 preconceptual design activities for European DEMO divertor have been conducted as an integrated, interdisciplinary R&D effort in the framework of EUROfusion Consortium. Consisting of two subproject areas, ‘Cassette’ and ‘Target’, this divertor project has the objective to deliver a holistic preconceptual design concept together with the key technological solutions to materialize the design. In this paper, a brief overview on the recent results from the subproject ‘Cassette’ is presented. In this subproject, the overall cassette system is engineered based on the load analysis and specification. The preliminary studies covered multi-physical analyses of neutronic, thermal, hydraulic, electromagnetic and structural loads. In this paper, focus is put on the neutronics, thermohydraulics and electromagnetic analysis.

The availability of a high flux neutron source for testing candidate materials under irradiation conditions, which will be typically encountered in future fusion power reactors (ITER, DEMO, FPR), is a fundamental step towards the development of fusion energy. To this purpose, the International Fusion Materials Irradiation Facility (IFMIF) represents the reference option to provide the fusion community with a DEMO-relevant neutron source capable of irradiating samples at a damage rate of up to 20 dpa/fpy (in steel) in a volume of 0.5 l. In the framework of the engineering design activities of IFMIF, ENEA is committed in the design of the lithium target assembly (TA) with removable (bayonet) backplate (BP) whose development has recently progressed up to a well advanced stage. However, an optimization of the system is still to be accomplished. In particular, the BP design needs to be revised in order to totally satisfy the ITER SDC-IC reference design criteria and fulfil the requirements on its lifetime which is limited by the neutron-induced swelling effects. In this work, a full thermo-mechanical analysis of the whole TA including a pseudo-transient simulation of the swelling effects in the BP over one year of full power operation has been performed by means of a 3D finite element (FE) model implemented through the quoted ABAQUS v. 6.14 code. A detailed neutronic analysis has been also performed by ENEA using the MCNP code to obtain the prompt nuclear responses, in terms of DPA and volumetric density of deposited nuclear heat power, to be used as input for the thermo-mechanical calculations. A new BP design capable to verify the SDC-IC design criteria and ensure its required swelling lifetime is proposed and described on the basis of the results of the performed analysis.

The EU-DEMO conceptual design is being conducted among research institutions and universities from 26 countries of European Union, Switzerland and Ukraine. Its mission is to realise electricity from nuclear fusion reaction by 2050. As DEMO has been conceived to deliver net electricity to the grid, the choice of the Breeding Blanket (BB) coolant plays a pivotal role in the reactor design having a strong influence on plant operation, safety and maintenance. In particular, due to the pulsed nature of the heat source, the Primary Heat Transfer System (PHTS) becomes a very important actor of the Balance of Plant (BoP) together with the Power Conversion System (PCS). Moreover, aiming to mitigate the potential negative impact of plasma pulsing on BoP equipment, for the DEMO plant is also being investigated a “heat transfer chain” option which envisages an Intermediate Heat Transfer System (IHTS) equipped with an Energy Storage System (ESS) between PHTS and PCS. Within this framework, a preliminary study has been carried out to analyse the thermal dynamic behaviour of the IHTS system for the Helium-Cooled Pebble Bed (HCPB) BB concept during pulse/dwell transition which should be still considered as the normal operating mode of a fusion power plant. Starting from preliminary thermal-hydraulic calculations made in order to size the main BoP components, the global performances of DEMO BoP have been quantitatively assessed focusing the attention on the attitude of the whole IHTS to smooth the sudden power variations which come from the plasma. The paper describes criteria and rationale followed to develop a numerical model which manages to simulate simple transient scenarios of DEMO BoP. Results of numerical simulations are presented and critically discussed in order to point out the main issues that DEMO BoP has to overcome to achieve a viable electricity power output.

One of the crucial steps toward the commercial exploitation of fusion energy is the availability of a high-intensity neutron source able to test and qualify structural materials to be used in future fusion power reactors. The International Fusion Materials Irradiation Facility-DEMO Oriented NEutron Source (IFMIF-DONES) facility, which is currently being designed within the framework of the EUROfusion Consortium, represents the European effort to develop such a neutron source. It employs a liquid lithium target struck by an accelerated deuteron (D+) beam to produce fusion-like neutrons used to irradiate materials samples. So far, a detailed investigation of the thermomechanical behavior of the IFMIF-DONES target system has been accomplished for the case of steady-state operation, considering the D+ beam steadily running at full nominal power (5 MW). However, the study of the effects generated in the lithium and in the surrounding steel structure backplate (BP) during the unavoidable beam trips needs to be carried out as well to exclude the possibility of system failure by fatigue phenomena. To this aim, a numerical assessment of the fatigue damage produced on the BP by repeated thermally induced shock waves propagating in the lithium as a consequence of accelerator fast beam trip events has been carried out following the methodology outlined in the RCC-MRx nuclear design code. In this article, the details of the calculations and the results obtained from such analysis are presented showing the capability of the system to survive for its whole expected lifetime.

The EU-DEMO Water-Cooled Lithium Lead Breeding Blanket (WCLL BB) concept foresees liquid Pb-15.7Li eutectic alloy as breeder and neutron multiplier, whereas pressurized subcooled water as coolant, with operative conditions typical of the PWR fission reactors (temperature in the range of 295–328 °C and pressure of 15.5 MPa). The cooling down of the BB is guaranteed by means of two separated cooling circuits: the one consisted in square channels housed within the complex of Side Walls and First Wall, and the one composed of a set of Double-Walled Tubes (DWTs) submerged in the Breeding Zone (BZ) and deputed to remove heat power therein generated. A parametric thermal study has been carried out in order to optimize the DWTs lay-out within the BZ of the WCLL BB outboard equatorial region ensuring that EUROFER components maximum temperature stays below the allowable value of 550 °C. The analysis has been performed following a theoretical-numerical approach based on the Finite Element Method (FEM) and adopting the ABAQUS qualified commercial code. In the paper, results obtained together with the main assumptions adopted in the parametric analysis are critically discussed and a new, sustainable DWTs configuration is proposed.

Within the framework of EUROfusion R&D activities, CEA-Saclay has carried out an investigation of the thermal and mechanical performances of alternative designs intended to enhance the Tritium Breeding Ratio (TBR) of the Helium-Cooled Lithium Lead (HCLL) Breeding Blanket (BB) for DEMO. Neutronic calculations performed on the 2014 DEMO HCLL baseline predicted a value of TBR equal to 1.07, lower than the required value of 1.1, necessary to ensure the tritium self-sufficiency of the breeding blanket taking into account uncertainties. In order to reach the TBR target, the strategy of the steel amount reduction inside the HCLL module breeding zone (BZ) has been followed by suppressing some stiffening/cooling plates inside the BZ, leading to this “advanced” concept. Since all the plates inside the BZ are actively cooled by helium, each change in their geometric layout has a strong impact on the thermal response of the module. Moreover, the removal of stiffening plate may impact the resistance of the box in case of in-module loss of coolant accident (LOCA). A thermal and mechanical campaign of analyses has been carried out in order to assess a potentially optimized layout of the module which could comply with the whole set of rules foreseen for the HCLL BB design. To perform this research campaign, a theoretical-numerical approach, based on the Finite Element Method (FEM), has been followed and the qualified Cast3M and NX FEM codes have been adopted. Results obtained are herewith reported and critically discussed, highlighting the open issues and suggesting the pertinent modifications to DEMO HCLL module design.

The design of the Demo Oriented NEutron Source (DONES) is the main objective of the Work Package Early Neutron Source (WPENS) of EUROfusion Power Plant Physics and Technology (PPPT) programme. DONES is an IFMIF-based neutron source with the goal of testing and qualifying candidate materials to be used in DEMO and future fusion power plants. In the framework of these activities, a research campaign has been carried out at the University of Palermo, in close cooperation with ENEA labs, in order to assess the thermomechanical performances of the DONES target system, endowed with an integrated Target Assembly (TA) when it undergoes the thermomechanical loads typical of its nominal operating scenario. The effects on TA behaviour of two possible different configurations for the accelerator beam foot-print have been investigated, considering an entire (20 × 5 cm2) or a halved (10 × 5 cm2) beam foot-print area with respect to that of IFMIF. The study has been carried out following a theoretical-numerical approach, based on the Finite Element Method (FEM), and adopting the qualified ABAQUS v. 6.14 commercial FEM code. The obtained thermomechanical results have been assessed in order to check whether the BP deformation field is such to prevent the contact of the BP itself with the High Flux Test Module, as well as to verify their compliance with the design criteria foreseen for the structural material. To this purpose, a stress linearization procedure has been performed along the most critical paths located within the BP, in order to check the fulfilment of the rules prescribed by the ITER structural design code SDC-IC. The obtained results are herewith presented and critically discussed.

Within the roadmap that will lead to the nuclear fusion exploitation for electric energy generation, the construction of a DEMOnstration (DEMO) reactor is, probably, the most important milestone to be reached since it will demonstrate the technological feasibility and economic competitiveness of an industrial-scale nuclear fusion reactor. In order to reach this goal, several European universities and research centres have joined their efforts in the EUROfusion action, funded by HORIZON 2020 UE programme. Within the framework of EUROfusion research activities, ENEA and University of Palermo are involved in the design of the Water-Cooled Lithium Lead Breeding Blanket (WCLL BB), that is one of the two BB concepts under consideration to be adopted in the DEMO reactor. It is mainly characterized by a liquid lithium-lead eutectic alloy acting as breeder (lithium) and neutron multiplier (lead), as well as by subcooled pressurized water as coolant. Two separate circuits, both characterized by a pressure of 15.5 MPa and inlet/outlet temperatures of 295 °C/328 °C, are deputed to cool down the First Wall (FW) and the Breeder Zone (BZ). The former consists in a system of radial-toroidal-radial C-shaped squared channels where countercurrent water flow occurs while the latter relies in the use of bundles of poloidal-radial Double Walled Tubes (DWTs) housed within the breeder. A parametric thermal study has been carried out in order to assess the best DWTs’ layout assuring that the structural material maximum temperature does not overcome the allowable limit of 550 °C and that the overall coolant thermal rise fulfils the design target value of 33 °C. The study has been performed following a theoretical-numerical approach based on the Finite Element Method (FEM) and adopting the quoted Abaqus FEM code. Main assumptions and models together with results obtained are herewith reported and critically discussed.

The European roadmap for the realization of fusion energy considers the stellarator line as a possible long-term alternative to a tokamak DEMO. In this context, from the plasma physics standpoint, the most promising option is a five-field period power plant called HELIcal-axis Advanced Stellarator (HELIAS) 5-B. In order to allow the electricity production, the HELIAS 5-B reactor must be endowed with a breeding blanket (BB). Hence, in this paper, the advancements in the HELIAS 5-B BB design are reported. In particular, the structural assessment of a whole BB period, extending along toroidal direction for 72 °, is depicted. A geometric configuration encompassing dummy BB segments has been set up to verify the feasibility of the proposed BB segmentation strategy for the whole sector. Both the Helium Cooled Pebble Bed (HCPB) and Water Cooled Lithium Lead (WCLL) BB concepts have been considered. Results have shown that, in principle, the occurrence of overlapping between adjacent segments depends on the BB-Vacuum Vessel connection mode, regardless of the considered BB concept, and on their mutual thermal expansion. Then, focusing only on the HCPB BB concept, a preliminary detailed geometric design of two BB segments, located at the inboard region of a bean shaped ring, has been carried out in order to investigate their structural behavior under steady state accidental loading conditions. To this end, the stress field arising within the structure of the two segments has been assessed following the RCC-MRx structural design rules. Results have shown that the proposed design works quite well under the assumed loading conditions. The study has been performed adopting the Finite Element Method and using the Ansys v.19.1 commercial code. The obtained results will be presented and critically discussed.

Within the framework of EUROfusion R&D activity, a research campaign has been carried out at the University of Palermo, in close cooperation with ENEA labs, in order to preliminary design the top cap foreseen for the DEMO Water-Cooled Lithium Lead (WCLL) breeding blanket segments. Due to the high heat and pressure loads acting on such component, its design results particularly demanding and a specific multi-physics approach is needed, covering several aspects from thermal-hydraulics to structural assessments. Preliminary detailed CAD model of the cap integrated into the upper region of the WCLL breeding blanket outboard central segment has been set-up, equipped with proper cooling circuits as well as manifold and attachment systems according to the design of the equatorial elementary cell. A detailed numerical model has been set-up with the aim of simulating the thermo-mechanical behaviour of the above-mentioned system. Finally, the thermo-mechanical response of the upper breeding blanket region has been evaluated in terms of stress and temperature distributions, verifying that the structural material maximum temperature stays below its limit value and that structural integrity is ensured by means of the fulfilment of design rules reported in RCC-MRx structural design code.