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This paper refers to the activity of structural design of DEMO Divertor in the framework of the EUROfusion Consortium. The structural analysis and its preparatory assessments were carried on since a year and the first results were published in a previous paper. The Cassette Body has been examined considering the most conservatives loads (e.g. coolant pressure, volumetric nuclear heating and electro-magnetic loads) according to their latest estimates. This work is based on the design-by-analysis approach adopted in the conceptual design phase of the DEMO Divertor. This design activity has been focused on some key parameters e.g. loads, main geometric dimensions, positions of the Cassette attachments on the vacuum vessel, way of loads application to characterize the structural behavior of the Divertor Cassette. In addition to the existing 3D solid element model, a shell element model has also been developed: with this new model a parametric analysis can be done for a fast optimization. The structural assessment was done according to the Design and Construction Rules for Mechanical Components of Nuclear Installation (RCC-MRx).

This paper describes the activity addressed to the conceptual design of the first wall and the main containment structures of DTT device. The work moved from the geometrical constraints imposed by the desired plasma shape and the configuration needed for the magnetic coils. Many other design constraints have been taken into account such as remote maintainability, space reservations for diagnostic and heating equipment, etc. The basic vessel design resulted in an all-welded single-wall toroidal structure made of 18 sectors. Proper supports have been designed for the first wall, which was conveniently segmented in view of remote maintenance. This provisional model allowed evaluating the electromagnetic loads on the metallic structure of the vacuum vessel resulting from the current quench due to a plasma disruption. After a FEA mechanical assessment, which was conducted according to ASME code, INCONEL® 625 has been provisionally selected as reference material for vacuum vessel. The design principles of the cryostat were chiefly based on cost minimization and functionality; thus it was conceived as a single-wall cylindrical vessel supported by a steel frame structure. The same structure will hold the vacuum vessel and the magnets.

This paper describes the development of a master model concept of the DEMO vacuum vessel (VV) conducted within the framework of the EUROfusion Consortium. Starting from the VV space envelope defined in the DEMO baseline design 2014, the layout of the VV structure was preliminarily defined according to the design criteria provided in RCC-MRx. A surface modelling technique was adopted and efficiently linked to the finite element (FE) code to simplify future FE analyses. In view of possible changes to shape and structure during the conceptual design activities, a parametric design approach allows incorporating modifications to the model efficiently.

One of the most critical components in the design of DEMO Power Plant is the Breeding Blanket (BB). Currently, four candidates are under study as options for DEMO. One of these is the Water Coolant Lithium Lead (WCLL) Breeding Blanket (BB). During the previous years a conceptual design of WCLL BB has been developed. At the current stage some open issues related to its manufacturability and design are still under evaluation. Since DEMO project is still in the pre-concept phase, the WCLL BB design team decided to investigate, in parallel with the current studies, an alternative design of the WCLL BB internal structure inspired to studies conducted in ‘90 s. The main drivers, in developing the alternative design, consisted on reducing the complexity of the internal structure itself and increasing the performances of the single module in terms of neutron shielding, tritium self-sufficiency, heat extraction and transportation to the primary heat transfer system and magneto hydrodynamic effects. All that, taking into account the lesson learned by the studies carried out in the last three years. The segment has been conceived as a single box where the lithium lead flows inside the module in poloidal direction. The rear part of the module is entirely dedicated to the cooling water manifold. A first 3d model of the alternative design has been developed, structural analyses have been carried out to optimize the internal structure against an over pressurization scenario. The results and the optimized design of the alternative WCLL BB design are here presented and discussed. © 2019 Elsevier B.V.

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.

One of the most critical components in the design of DEMO Power Plant is the Breeding Blanket. Currently, four candidates are investigated as options for DEMO. One of these is the Water Coolant Lithium Lead Breeding Blanket. A new concept design has been proposed and investigated in 2015. The first activity driving the Breeding Blanket design was the definition of the poloidal segmentation. Current trend in breeding blanket design is based on the multi module box approach, which has advantages in terms of manufacturing, in reducing the global stress and strain during the start-up and the shut-down phases and during operation, because the favourable thermal expansions; and in simplifying the First Wall layout and integration. Nevertheless, drawbacks are identified, such as the reduction of Tritium Breeding Ratio, the constraints in manifold and in Back Supporting Structure design and integration because the limited space available. The present work concerns a method that, starting from these constraints, defines and optimizes some of the main design drivers for the selection of the segmentation of the Water Coolant Lithium Lead Breeding Blanket. The method by definition is based just on geometrical parameters because it is used as first step of the design when any analysis and detailed data are available. It is based on the definition of Figures Of Merits, consisting in numerical parameters, such as the ratio between the modules volume and the overall volume of segment assigned, the approximation between the real profile of the modules and the theoretical one, the form factor of the modules, the ratio between the module thickness at the mid-plane and the segment thickness at the same position. The figures of merits support the choice among different options. In particular two different solutions of poloidal segmentation have been compared and, according to the proposed method, the best one was chosen for the design of Water Coolant Lithium Lead Breeding Blanket.

The EUROfusion Consortium is setting up – as part of the EU Fusion Roadmap – the framework for the implementation of the (pre)conceptual design phase of the DEMO reactor. Configuration management needs have been identified as one of the key elements of this framework and is the topic of this paper, in particular the configuration of the CAD design data. The desire is to keep the definition and layout of the corresponding systems “light weight” and relatively easy to manage, whilst simultaneously providing a level of detail in the definition of the design configuration that is fit for the purpose of a conceptual design. This paper aims to describe the steps followed during the definition of the configuration management system of the DEMO design data in terms of (i) the identification of the appropriate product data management system, (ii) the description of the philosophy of the configuration management of the design data, and (iii) the introduction of the most important enabling processes.

The present work focuses on the design assessment of the DEMO Upper Port. The size of the upper port is defined by the available space in between the toroidal field coils and the required space to integrate a thermal shield between the vacuum vessel (VV) port and the coils. Since the large breeding blanket (BB) segments will require periodic replacement via the upper vertical ports the space inside the upper port needs to be maximized. For this reason the optimization and verification of the upper port design is a critical aspect in the development of DEMO project. The work here presented investigates the possibility to have an upper port with single walled sidewalls to increase the space available inside the port for the integration of pipe work and to allow the handling of the BB segments. The work carried out evaluates the feasibility of the design solution from the structural and thermal point of view verifying the upper port structure against nuclear heating, in-vessel pressure, and electromagnetic loads due to a toroidal field coil fast discharge and plasma disruptions according to nuclear codes. © 2018 Elsevier B.V.

The objective of the work is to describe the design and the realization of a virtual simulator of a metropolitan railway cockpit, aimed at improving the perception of safety by means of tests made by users in Virtual Reality, analysed through statistical methodologies. The user lives the experience of a driver in an immersive and interactive Augmented Reality session, interacting with the train dashboard and all its control and signalling devices. In particular, the user is proposed to test different dashboards, different configurations of the controls and different signalling and safety devices in order to compare different concept and select the optimum in terms of perception of dangerous situation, reaction to an event and cognitive response in different situations of the rail vehicle driving. The simulator consists of a simulacrum integrating different technologies, physically composed of a dashboard of the cockpit of a metropolitan train and a real seat. The geometry of the dashboard has been acquired through Reverse Engineering techniques from a real train dashboard. The user’s immersion in the virtual environment during the simulation is guaranteed by the scene displayed on the Augmented Reality device, while, simultaneously, the stereoscopic projection on a screen above the dashboard makes available the experience even to users not directly involved, seeing the scene from the driver’s point of view. The immersive Augmented Reality is realized through a Head-Mounted Display (HMD) by which the user, protagonist of the driving experience, sees the configuration of the virtual control devices (CAD geometries) overlapped with the physical dashboard in order to naturally interact into the immersive environment. The interaction between user and simulator happens through the NUI (Natural User Interfaces) based on markerless tracking of parts of the user’s body.

This paper deals with pre-concept studies of DEMO divertor cassette-to-vacuum vessel locking system under the work program WP13-DAS-07-T06: Divertor Remote Maintenance System pre-concept study. An iterative design process, consistent with Systems Engineering guidelines and named Iterative and Participative Axiomatic Design Process (IPADeP), is used in this paper to propose new innovative solutions for divertor locking system, which can overcome the difficulties in applying the ITER principles to DEMO. The solutions conceived have been analysed from the structural point of view using the software Ansys and, eventually, evaluated using the methodology known as Fuzzy-Analytic Hierarchy Process. Due to the lack and the uncertainty of the requirements in this early conceptual design stage, the aim is to cover a first iteration of an iterative and incremental process to propose an innovative design concept to be developed in more details as the information will be completed.

This paper focuses on a preliminary structural analysis of the current concept design of DEMO vacuum vessel (VV). The VV structure is checked against a vertical load due to a Vertical Displacement Event in combination with the weight force of all components that the main vessel shall bear. Different configurations for the supports are considered. Results show that the greatest safety margins are reached when the tokamak is supported through the lower ports rather than the equatorial ports, though all analyzed configurations are compliant with RCC-MRx design rules.

The present work focuses on structural assessment of DEMO Vacuum Vessel Lower Port structure. Since previous studies have been addressed the structural scheme of the main vessel, this work investigates a feasible layout of vessel supports defining the position of the pumping port cut and different inclinations of the lower port. All design configurations have been analysed according to Design and Construction Rules for Mechanical Components of Nuclear Installations. The structure was checked against a vertical load due to a Vertical Displacement Event in combination with the estimated mass of all components supported by the vessel. The outcome of the assessment gives relevant information about the optimal position of the supports, the impact of the pumping port duct cut and the lower port inclination.

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.

In the EU DEMO design (Romanelli, 2012; Federici et al., 2014), due to the large number of complex systems inside the tokamak vessel it is of vital importance to address the in-vessel integration at an early stage in the design process. In the EU DEMO design, after a first phase in which the different systems have been developed independently based on the defined baseline DEMO configuration, an effort has been made to define the interface requirements and to propose the strategies for the mechanical integration of the auxiliary heating and fuelling systems into the Vacuum Vessel and the Breeding Blanket. This work presents the options studied, the engineering solutions proposed, and the issues highlighted for the mechanical in-vessel integration of the DEMO fuelling lines, auxiliaries heating systems, and diagnostics. (C) 2017 The Author(s). Published by Elsevier B.V.

This paper presents the pre-conceptual design activities conducted for the European DEMO divertor, focusing on cassette design and Plasma Facing Components (PFC) integration. Following the systems engineering principles, a systematic design method, the Iterative and Participative Axiomatic Design Process (IPADeP), has been adopted. Basing on Axiomatic Design, IPADeP supports the early conceptual design of complex systems. The work moved from the geometrical and interface constraints imposed by the 2015 DEMO configuration model. Then, since different materials will be used for cassette and PFCs, the divertor geometry has been developed taking into account the cooling parameters of the cassette Eurofer steel and the integration of PFCs cooling system. Accordingly, the design process led to a double wall cassette structure with internal reinforcing ribs to withstand cassette coolant pressure and three different kinds of piping schemes for PFCs with dual circuits. These three solutions differs in the feeding pipes layouts and target manifold protection and they have been proposed and evaluated considering heat flux issues, shielding problems, interface requirements with blanket and vacuum vessel and remote maintenance needs. A cassette parametric shell model has been used to perform first structural analyses of the cassette body against coolant pressure. Taking advantages of the parametric surface modelling and its linkage with Finite Element (FE) code, the cassette ribs layout and thickness has been evaluated and optimized, considering at the same time the structural strength needed to withstand the coolant parameters and the maximum stiffness required for cassette preloading and locking needs.

This paper is based on the remote maintenance system project (WPRM) for the demonstration fusion power reactor (DEMO). Following ITER, DEMO aims to confirm the capability of generating several hundred of MW of net electricity by 2050. The main objective of these activities is to develop an efficient and reliable remote handling (RH) system for replacing the divertor cassettes. This paper presents the preliminary results of the concept design of the divertor RH system. The proposed divertor mover is a hydraulic telescopic boom driven from the transportation cask through the maintenance tunnel of the reactor. The boom is divided in three sections of 4 m each, and it is driving an end-effector in order to perform the scheduled operations of maintenance inside the vacuum vessel. Two alternative design of the end effector to grip and manipulate the divertor cassette are also presented in this work. Both the concepts are hydraulically actuated, basing on the ITER previous studies. The divertor cassette end-effector consists of a lifting arm linked to the divertor mover, a tilting plate, a cantilever arm and a hook-plate. The main objective of this paper is to illustrate the feasibility of DEMO divertor remote maintenance operations.

Fusion advanced study torus (FAST) has been designed with the aim to tackle the power exhaust problem with ITER and DEMO relevant bulk plasma. Considerable analysis effort has been spent in EM (electromagnetic) designing of FAST divertor components capable of withstanding the electromagnetic loads expected from the foreseen FAST operative conditions. Plasma disruption EM loads are one of the most critical load conditions for the divertor. Consequently a first dimensioning of the divertor for EM loads is mandatory. The foreseen FAST divertor structure is compact and the configuration is aimed to give to the structure the required mechanical and thermal capability as well as being remote handling compatible. The conceptual EM design of the divertor has been designed for FAST operative conditions and for DEMO relevant "Snow Flake" (SF) magnetic topology. Recently a "Snow Flake" (SF) magnetic topology has been suggested for the divertor region, capable to spread the plasma power flow on a much wider areas, with the possibility of reducing by a factor 4 the power flowing to the divertor tiles [1]. The conceptual design of the divertor has been obtained during activities based on the estimation of EM loads due to a typical Plasma Fast Down disruption event in FAST for the normal configuration and for SF magnetic topology. © 2013 Elsevier B.V.

The paper presents a concept design of a remote handling (RH) system oriented to maintenance operations on the divertor second cassette in FAST, a satellite of ITER tokamak. Starting from ITER configuration, a suitably scaled system, composed by a cassette multifunctional mover (CMM) connected to a second cassette end-effector (SCEE), can represent a very efficient solution for FAST machine. The presence of a further system able to open the divertor port, used for RH aims, and remove the first cassette, already aligned with the radial direction of the port, is presumed. Although an ITER-like system maintains essentially shape and proportions of its reference configuration, an appropriate arrangement with FAST environment is needed, taking into account new requirements due to different dimensions, weights and geometries. The use of virtual prototyping and the possibility to involve a great number of persons, not only mechanical designers but also physicist, plasma experts and personnel assigned to remote handling operations, made them to share the multiphysics design experience, according to a concurrent engineering approach. Nevertheless, according to the main objective of any satellite tokamak, such an approach benefits the study of enhancements to ITER RH system and the exploration of alternative solutions.

The Fusion Advanced Study Torus (FAST), with its compact Tokamak design, high toroidal field and plasma current, will face many of the problems that ITER will meet and will anticipate many DEMO relevant physics and technology issues. The Design Upgrade of the Vessel and In-Vessel Components is presented in this paper. Relevant modifications were performed to the Vacuum Vessel (VV) and to the Plasma Facing Components (PFCs), i.e. the First Wall (FW) and the Divertor. The VV was modified to insert active reduction coils (ARC), between VV and the toroidal field (TF) coils to keep toroidal field magnet ripple lower than 0.3% and to allow Remote Handling for the FW and the Divertor. The FW, was modified to house coils for ELMs control and other plasma instabilities. A 3D thermo-hydraulic analysis using ANSYS code was performed to check FW heat removal capability. A new Divertor was designed to withstand the largest thermal loads of the high performance, low density, H-mode and to be able to comply with a recent magnetic topology called as "Snow Flake", increasing up a factor 4 the flux expansion. An exhaustive 3D thermo-hydraulic analysis using ANSYS code was carried out to show the capability of the Divertor to comply these high requirements. Design criteria were satisfied by present components of the upgraded machine. © 2013 Elsevier B.V.

One of the fundamental input parameters required for the thermo hydraulic and structural design of a divertor cassette is the operation temperature range. In the current design activities to develop European DEMO divertor in the frame of EUROfusion, reduced activation steel EUROFER97 was chosen as structural material for the divertor cassette body considering its low long-term activation and superior creep and swelling resistance under neutron irradiation (You et al., 2016) [1] . For specifying an operation temperature range (i.e. cooling condition) various, often conflicting requirements have to be considered. In this article the lower limit of allowed operation temperature window is defined for EUROFER97 for structural design of DEMO divertor cassette body. The underlying rationale and supporting experimental data from a number of previous irradiation tests are also presented. The motivation of this survey study is to explore the possibility to use EUROFER97 for water-cooled divertor cassette at temperatures below 350 °C which has been regarded as limit temperature to preserve ductility under irradiation. Based on the literature data of FTTT (Fracture Toughness Transition Temperature) calibrated by Master Curve method, it is concluded that EUROFER97 at the envisaged maximum dose of 6 dpa will have to be operated above 180 °C taking the embrittlement due to helium production into account.

This paper explores a possible integration of some ancillary systems of helium-cooled lithium lead (HCLL) and helium-cooled pebble-bed (HCPB) test blanket modules in ITER CVCS area. Computer-aided design and ergonomics simulation tools have been fundamental not only to define suitable routes for pipes, but also to quickly check for maintainability of equipment and in-line components. In particular, accessibility of equipment and systems has been investigated from the very first stages of the design using digital human models. In some cases, the digital simulations have resulted in changes in the initial space reservations. © 2015 Elsevier B.V.

The WEST project (W-Environment in Steady-State Tokamak) aims to transform the Tore Supra limiter configuration to an x-point divertor, providing a test bed for ITER-like plasma-facing components (actively cooled W monoblocs) under high heat flux, steady-state plasma irradiation. The lower divertor includes an actively cooled, W-coated CuCrZr baffle to provide neutral compression and improve particle exhaust. As part of the new diagnostic equipment of Tore Supra within WEST project, a set of Langmuir probes will find place on the baffle in order to provide plasma flux and electron temperature measurements for physics studies and real-time machine protection functions during steady-state discharges. On the baffle top surface, irradiation coming from the plasma, energetic ripple-ions losses, photons and energetic neutrals from charge exchange reactions produce power fluxes up to 3 MW/m2, representing a challenge for the Langmuir probes operating conditions. In this paper Copper–Chrome–Zirconium (CuCrZr) cylindrical probe concept design is proposed. Finite element thermo-mechanical analysis (FEA) confirmed the consistency of this solution under the steady-state plasma condition in the worst case (highest thermal load).

Divertor is a crucial component in Tokamaks, aiming to exhaust the heat power and particles fluxes coming from the plasma during discharges. This paper focuses on the optimization process of FAST divertor, aimed at achieving required thermo-mechanical capabilities and the remote handling (RH) compatibility. Divertor RH system final layout has been chosen between different concept solutions proposed and analyzed within the principles of Theory of Inventive Problem Solving (TRIZ). The design was aided by kinematic simulations performed using Digital Mock-Up capabilities of Catia software. Considerable electromagnetic (EM) analysis efforts and top-down CAD approach enabled the design of a final and consistent concept, starting from a very first dimensioning for EM loads. In the final version here presented, the divertor cassette supports a set of tungsten (W) actively cooled tiles which compose the inner and outer vertical targets, facing the plasma and exhausting the main part of heat flux. W-tiles are assembled together considering a minimum gap tolerance (0.1-0.5 mm) to be mandatorily respected. Cooling channels have been re-dimensioned to optimize the geometry and the layout of coolant volume inside the cassette has been modified as well to enhance the general efficiency. © 2015 Elsevier B.V. All rights reserved.

The fusion advanced study torus (FAST), with its compact design, high toroidal field and plasma current, faces many of the problems met by ITER, and at the same time anticipates much of the DEMO relevant physics and technology. The conceptual design of the first wall (FW) and the vacuum vessel (VV) has been defined on the basis of FAST operative conditions and of "Snow Flakes" (SF) magnetic topology, which is also relevant for DEMO. The EM loads are one of the most critical load components for the FW and the VV during plasma disruptions and a first dimensioning of these components for such loads is mandatory. During this first phase of R&D activities the conceptual design of the FW and VV have been assessed estimating, by means of FE simulations, the EM loads due to a typical vertical disruption event (VDE) in FAST. EM loads were then transferred on a FE mechanical model of the FAST structures and the mechanical response of the FW and VV design for the analyzed VDE event was assessed. The results indicate that design criteria are not fully satisfied by the current drawing of the VV and FW components. The most critical regions have been individuated and the effect of some geometrical and material changes has been checked in order to improve the structure. © 2015 Elsevier B.V. All rights reserved.

In view of the ITER conceptual design review, the design of the ancillary systems of the European test blanket systems presented in [1] has been updated and made consistent with the ITER requirements for the present design phase. Europe is developing two concepts of TBM, the helium cooled lithium lead (HCLL) and the helium cooled pebble bed (HCPB) one, having in common the cooling media, pressurized helium at 8 MPa [2]. TBS, namely helium cooling system (HCS), coolant purification system (CPS), lead lithium loop and tritium extraction/removal system (TES-TRS) have the purpose to cool down the TBM and to remove tritium to be driven to TEP from breeder and coolant. These systems are placed in port cell 16 (PC#16), chemical and volume control system (CVCS) area and tritium building. Starting from the pre-conceptual design developed in the past, more mature technical interfaces with the ITER facility have been consolidated and iterative design activities were performed to comply with design requirements/specifications requested by IO to conclude the conceptual design phase. In this paper the present status of design of the TBS is presented together with the preliminary integration in ITER areas. © 2015 Elsevier B.V.

The Fusion Advanced Study Torus (FAST) has been proposed as a high magnetic field, compact size tokamak providing a flexible integrated environment to study physics and technology issues in ITER and DEMO relevant conditions. FAST has a quite large natural toroidal field ripple (around 1.5%) due to its compactness and to the number of access ports: this ripple must be lowered to an acceptable level to allow safe operations and a good confinement quality. An Active Ripple Compensating System (ARCS) has been designed, based on a set of poloidal coils placed between the plasma chamber and the Toroidal Field Coils (TFCs). These ARCS coils will be fed with adjustable currents, opposite in direction respect to the TFC currents, and will allow lowering the ripple up to zero and beyond. The CAD model of FAST including the ARCS coils has been completed and preliminary electromagnetic and thermal analyses have been carried out. Moreover, a Feedback Active Control System (FACS) composed of two arrays of in-vessel saddle coils has been designed to allow safe high plasma current, low safety factor operation and to mitigate possibly large ELMs effects in FAST. These FACS coils will be fed by a feedback system to control MHD modes: a first engineering assessment of the current requirements has been carried out. © 2013 Euratom-ENEA Association sulla Fusione.

This paper presents the recent progress in the pre-conceptual design activities for the EU-DEMO divertor Cassette Body, performed in the framework of the work package “Divertor” of the EUROfusion Power Plant Physics & Technology (PPPT) program. According to Systems Engineering Principles, the divertor CAD model is reviewed, considering the updates in the DEMO configuration model presented by the Programme Management Unit (PMU) in 2017. The design parameters affected by these changes and their impact on the divertor design and on the interfaced systems are analysed. Then, the paper focuses on the integration on the new cassette geometry of the divertor sub-systems. This includes the design of a “shielding liner” for cassette body and Vacuum Vessel protection, as well as the development of the cassette body-to-Vacuum Vessel fixation system. The design activities related to these main sub-systems are discussed in detail, in terms of CAD model and thermo-mechanical calculations.