Search

One of the most challenging objectives of the European research concerning nuclear fusion technology, promoted by the EUROfusion consortium, is to bring stellarator-type nuclear fusion devices to maturity. To this purpose, studies on a large HELIcal-axis advanced stellarator (HELIAS), extrapolated from Wendelstein 7-X and based on a 5-fold symmetry (HELIAS 5-B), are currently ongoing. The HELIAS 5-B stellarator reactor will be endowed with a breeding blanket (BB) system to allow for the self-sustainability of the nuclear fusion reaction and make it suitable for electricity generation. In this paper, we present the first ever heterogeneous mechanical design and the preliminary structural assessment of a bean-shaped ring of a HELIAS 5-B BB sector. The proposed mechanical design, which is based on the helium-cooled pebble bed (HCPB) BB concept and developed according to the “sandwich” architecture, foresees an actively cooled segment box connected to a back-supporting structure equipped with manifolds. The internal region (breeding zone) is reinforced by actively cooled steel plates. The proposed heterogeneous design was checked against nominal loads and an in-box loss of coolant accidental scenario, which is a typical design driver for BBs. The assessment has been performed according to the RCC-MRx structural design code. Our results are herewith presented and critically discussed, focusing on the potential follow-up of the HELIAS 5-B HCPB BB design.

The optimization of the aeronautical structure manufacturing is a challenging task in the development of a new aircraft. To date, aeronautical industries are funding research about new assembly approaches based on cost reduction and increased efficiency of the assembly processes. The work here presented focused on an innovative assembly method based on the integration between statistical methods of tolerance prediction and the determinant assembly approach. The coupling tolerances between airframe components are predicted through statistical approach in order to reduce the features manufactured in assembly. This aspect contributes to a reduction of the costs due non-recurring costs. The method proposed has been tested on a dedicated case study developed in the frame of the “integrated main landing gear box” project on the CleanSky2 Research program. Tests have been conducted to check the consistency of the method and its feasibility in the industrial contexts in the case of flat-shaped component. The performed experiments confirmed the analytical study.

The realization of nuclear fusion reaction as energy source is under investigation, among the scientific community, through the design and development of tokamak reactors. Among the several experiments worldwide, the ITER project is the ma-jor international experiment and it involves several research institutes from sev-eral countries. In such a project, a Systems Engineering (SE) approach is re-quested to organize and manage the design due to its highly integrated design, the safety requirements related to nuclear aspects and the complex procurement scheme. The SE discipline focuses the attention on the requirements which are crucial for every successful project, defining what the stakeholders want from a potential new system, namely what the system must do to satisfy stakeholders need. Correctly stating WHAT is needed for the system, it is possible to obtain its conceptual design (HOW) as much as possible complying the requirements. The incorrect definition of requirements often leads to the failing of a project. Stakeholders’ needs are written in Natural Language that is generally ambiguous, imprecise, incomplete and redundant. Their transformation into SMART require-ments is crucial to avoid design failure. However, it requires a great expertise, unless a specific procedure is assessed. To this end, this work presents a specific procedure based on “like-mind” processes to make systematic the SMART re-quirements definition and assessment from stakeholders needs. The procedure is based on a demand/response framework and it is developed to obtain ITER re-quirements. However, it can be easily extended to every project using its own specifications. A specific case study on ITER Remote Handling is presented in this paper as example of the conceived requirements transformation procedure.

Today it is more and more mandatory for all commercial companies to comply with the principles and methodologies of Industry 4.0 and to achieve the related capabilities protecting their competitiveness and taking a leading-edge position on market as regards technologies. Specifically, the whole production and sale system must achieve the fundamental characteristics of Industry 4.0 approach, but specially the manufacturing companies must also change and update their management procedures, internal organization, resource training, assets and all production process to keep safe their current business capacities. This evolution process is even more critical for Small and Medium Enterprises (SME), that traditionally tend to be conservative and to protect their way of operation, usually characterized by a low level of automation. The work presented focuses on the design and integration of a semi-automatic welding cell of train bolster in a SME which is currently realizing a project aimed to the acquisition of Industry 4.0 capabilities, with special focus on manufacturing processes. Among them, one of the most important is the production of welded-steel critical structures, that the Company supplies to prime manufacturer of railway rolling stock systems. The experience gained during the activity, the criticalities due to the integration processes and the adopted design methodologies are here described. The work has been carried out consistently with the Systems Engineering principles, starting from the requirements elicitation and analysis to the systematic approach for the design and integration activities.

Pellet injection represents, to date, the most promising option for core fuelling of the EU-DEMO tokamak. Simulations with the HPI2 pellet ablation/deposition code indicate, however, that sufficiently deep fuel deposition requires injection from the High Field Side (HFS) at velocities ≳1 km/s. Two complementary inboard injection schemes are being explored: one makes use of guide tubes with curvature radii ≥6 m in the attempt of preserving pellet integrity at speeds of ˜1 km/s, the other is investigating the feasibility of injecting high-speed (˜3 km/s) pellets along “direct line of sight” (DLS) trajectories, from either the HFS or a vertical port. Options using quasi-vertical DLS paths routed across the upper vertical port have been explored first, as they can be more easily integrated, Unfortunately, the radial position of the available vertical access (≳9 m from the machine axis) turns out to be unfavorable; further simulations with the HPI2 code predict indeed that vertical injection may be effective only if pellets trajectories are well inboard the magnetic axis. High-speed injection through oblique inboard “DLS” paths, not interfering with the Central Solenoid (CS), are instead predicted to yield good performance, provided that the injection location is ≲2.5 m from the equatorial mid-plane. The angular spread of high-speed free-flight pellets, recently measured using an existing facility, turns out to be enclosed within ˜ 0.7°. This scatter cone may require significant cut off volume of the Breeding Blanket (BB). Moreover, DLS in-vessel conical penetrations may increase the neutron flux outside of the bio-shield, and also result in a significant heat load in the cryogenic pellet source. These issues are being investigated, to identify suitable shielding strategies; preliminary results are reported. The suitability of straight guide tubes to reduce the scatter cone, and hence the corresponding open cross section on BB penetration and the neutron streaming, will be explored as a further step. © 2019 Elsevier B.V.

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.

In the current pre-concept phase of the European DEMO, integration studies of the systems in the Upper Port area are being carried out. In DEMO, the Upper Port of the Vacuum Vessel is extraordinarily large to allow for the vertical extraction of the Breeding Blanket segments. This requires a number of components inside and outside the port to be integrated with tight space constraints: The Upper Port structure and its annexes, the adjacent Toroidal and Poloidal Field Coils, the Thermal Shields, the piping connection to the Vacuum Vessel Pressure Suppression System, the Shield Plug and its inserts, the feeding pipework of the in-vessel components and part of the Breeding Blanket supporting structures. Apart from functional aspects, the design of these components is driven by considerations of structural integrity, maintainability and irradiation shielding, which are mutually competing in many areas. Several studies were conducted on the design of the Upper Port and the required configuration of the components within. The present article describes the development approach, the studied options and the respective results, the identified issues as well as the proposed engineering solutions, in particular with respect to the mechanical design of the Upper Port and the integrated Shield Plug.

In this work we present the latest progresses (September 2018) in the conceptual design of the main containment structures of DTT fusion reactor. The previous DTT baseline design is revised in terms of structural materials and overall reactor shape. The major change involves the vacuum vessel, which now foresees a welded double-wall stainless steel structure. The basic design includes eighteen sectors, with novel ports configuration for remote maintenance systems, diagnostics and heating equipment. New supports are designed for the first wall, which is conveniently segmented in view of assembly and remote replacement. The cryostat of the machine is conceived as a single-wall cylindrical vessel reinforced by ribs. The cryostat base is also in charge of supporting the vacuum vessel and the magnets system. A preliminary FEA analysis confirms that the main mechanical structure might withstand the design loads, in particular the ones resulting from possible plasma disruptions.

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 EUROfusion Consortium programme, the Water Cooled Lithium Lead (WCLL) option has been chosen as a candidate for the breeding blanket (BB) of the European fusion power demonstration plant (DEMO) conceptual design. Neutronic analyses play a fundamental role in the development of the WCLL blanket, providing guidelines for its design based on the evaluation of the nuclear performances. A detailed three-dimensional MCNP model of the latest WCLL layout has been generated and integrated in a DEMO MCNP generic model suitably designed for neutronic analyses. Three-dimensional neutron and gamma transport simulations have been performed using the MCNP5v1.6 Monte Carlo code and JEFF 3.2 nuclear data libraries, in order to assess the WCLL-DEMO tritium self-sufficiency and the shielding capabilities of the breeding blanket/manifold system to protect the vacuum vessel and toroidal field coils. Furthermore, radial profiles of the neutron flux, nuclear heating, neutron damage and he-production have been assessed in the inboard and outboard equatorial planes. The outcome of the present study highlights the potential and suitability of the WCLL breeding blanket for the application to DEMO, both in terms of tritium production and shielding performances.

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.

Summary The water-cooled lithium–lead breeding blanket is a candidate option for the European Demonstration Power Plant (DEMO) nuclear fusion reactor. This breeding blanket concept relies on the liquid lithium–lead as breeder–multiplier, pressurized water as coolant, and EUROFER as structural material. The current design is based on DEMO 2015 specifications and represents the follow-up of the design developed in 2015. The single-module-segment approach is employed. This is constituted by a basic geometry repeated along the poloidal direction. The power is removed by means of radial–toroidal (i.e., horizontal) water cooling tubes in the breeding zone. The lithium–lead flows in a radial–poloidal direction. On the back of the segment, a 100-mm-thick plate is in charge of withstanding the loads due to normal operation and selected postulated initiating events. 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 the single-module-segment water-cooled lithium–lead breeding blanket design are discussed and supported by thermo-mechanic, thermo-hydraulic, and neutronic analyses. Preliminary integration with the primary heat transfer system, the energy storage system, and the balance of plant is briefly discussed. Open issues, areas of research, and development needs are finally pointed out. @EUROfusion Consortium*, 2017. *For more details see http://www.euro-fusionscipub.org/disclaimer-copyright

This research discusses the use of a systematic design method, the Iterative and Participative Axiomatic Design Process (IPADeP), for the early conceptual design stage of large-scale engineering systems. The involvement of multiple and competing requirements has imposed high challenges for achieving an affordable design of complex systems in a reasonable lead time. Systems Engineering (SE) focuses on how to design and manage complex systems over their life cycles. Both must begin by discovering the real problems that need to be resolved and identifying from the early stage of the design the main stakeholder requirements and customer needs. The Axiomatic Design (AD) methodology is widely recognized in the literature to efficiently support the design of complex systems from the early conceptual stage. IPADeP provides a systematic methodology for applying AD theory in the conceptual design of large-scale engineering systems, aiming to minimize the risks related to the uncertainty and incompleteness of requirements and to improve the collaboration of multi-disciplinary design teams. IPADeP has been adopted as design methodology in the pre-conceptual design stage of a subsystem of the DEMOnstration fusion power plant (DEMO): the divertor cassette body-to-vacuum vessel locking system. In this paper improvements in IPADeP are presented and its validity is discussed by presenting the application to the divertor system design.

In the frame of the EUROfusion roadmap, the development of a conceptual design for the Demonstration Fusion Power Reactor (DEMO), beyond ITER, is a key issue. The DEMO reactor shall guarantee the tritium self-sufficiency, generate electricity and operate as a test facility for the fusion power plant relevant technologies, such as the breeding blanket (BB). The Water Cooled Lithium Lead (WCLL) concept has been chosen as a candidate for the DEMO BB: it relies on liquid Lithium Lead as breeder and neutron multiplier, Eurofer as structural material and pressurized water as coolant. A detailed MCNP model of the latest WCLL BB layout has been generated and integrated in the DEMO MCNP generic model suitably designed for neutronic analyses. Three-dimensional neutron and gamma transport simulations have been carried out by means the MCNP Monte Carlo code and JEFF nuclear data libraries in order to assess the WCLL-DEMO performances in terms of tritium self-sufficiency and shielding effectiveness to protect the vacuum vessel and the toroidal field coils. Moreover, the impact on the Tritium production of the water content in the first wall (FW) and the effect of its pressure/temperature has been addressed. The outcomes of the present study provide guidelines for the optimization of the WCLL DEMO reactor nuclear performances through the assessment of the loads on sensitive components and the estimation of its potential tritium generation capabilities.

The pre-concept design of the DEMO Vacuum Vessel is going on in view of the 2020 gate review; moreover the nuclear heat loads on the vessel inner shell were determined and found to be about one order of magnitude higher compared to ITER. A subsequent thermal-structural analysis of the vessel inner shell revealed high thermal stresses and a large temperature gradient through the inner shell thickness. Given the simultaneous occurrence of primary membrane stresses in the entire vessel inboard wall and, in proximity of the vessel ribs, high bending stresses due to the coolant pressure, a survey of all options within the design rules was required to identify the inter-dependencies of the individual stress limits (primary membrane, primary bending, thermal membrane plus bending). In order to face this kind of issues a detailed assessment on the design of the inboard wall of DEMO Vacuum Vessel has been conducted and is presented here. The current work evaluates both P and S type damages for the inboard wall of DEMO Vacuum Vessel in case of high nuclear heat load, vacuum vessel coolant pressure and toroidal field coil fast discharge. The elastic analysis method has been used to check the rules for prevention of both types of damage. The rules applied to prevent the aforementioned damages are compliant to Level A criteria, in case of negligible creep and negligible irradiation. In order to check the structural integrity of the inboard wall of DEMO VV against high thermal and mechanical loads, optimization structural analyses were performed and checked against the rules provided in the applicable design code (RCC MRx).

The optimization of the aeronautical structures manufacturing is one of the most challenging tasks in development of a new aircraft. Nowadays, aeronautical industries are supporting researches deal with the development of new assembly approaches which aim at increasing efficiency and reducing cost of the processes. The work here presented focused on definition of a Jig-less assembly procedure of the “Integrated Main Landing Gearbox” (ITEM B). The project aims to develop a new generation of Lower Center Fuselage with an innovative landing system integrated in the fuselage itself. The ambition of ITEMB is the creation of a single integrated structure in composite material of the gear bay that reduces assembly costs, optimizing and integrating the entire design, construction and maintenance of the aircraft. The approach here described is based on the integration of different engineering disciplines, as such as the tolerance statistical prediction, the ergonomics, digital human modeling, manufacturing and measurement technologies. In particular, through an appropriate ergonomic analysis, an innovative assembly process of the gear bay was developed and optimized in a virtual environment pointing out the advantages and disadvantages with respect to a traditional assembly cycle. The assembly process is based also on the implementation of the assembly tolerance prediction.

The development of a conceptual design for the Demonstration Fusion Power Reactor (DEMO) is a key issue within the EUROfusion roadmap. The DEMO reactor is designed to produce a fusion power of about 2 GW and generate a substantial amount of electricity, relying on a closed tritium fuel cycle: it implies that the breeding blanket (BB) shall guarantee a suitable tritium production to enable a continuous operation without any external supply. The Water-Cooled Lithium Lead (WCLL) concept is a candidate for the DEMO BB: it uses liquid Lithium Lead as breeder and neutron multiplier and water in PWR condition as coolant. The neutronics analyses carried out in the past have been performed using a semi-heterogeneous representation of the BB, since the complexity of its structure makes the generation of a detailed MCNP model a very demanding and challenging task. Results highlighted good performances for the WCLL BB, both in terms of shielding effectiveness and tritium self-sufficiency. A recently updated assessment of the tritium breeding ratio (TBR) requirement for DEMO, considering margins for calculation uncertainties and incomplete models of the whole machine, led to the definition of a tentative 1.15 value for the TBR. Moreover, the implementation of an accurate BB neutronics model, consistent with the engineering design, is recommended for the evaluation of the tritium self-sufficiency. In order to tackle these issues, an MCNP model of the DEMO tokamak, integrating a fully heterogeneous WCLL BB has been developed for the first time, including an accurate description of the FW water channels, as well as a comprehensive definition of the breeding zone inner structure. A complete assessment of the WCLL BB nuclear performances, through 3D neutron and gamma transport simulations, has been carried out by means the MCNP Monte Carlo code and JEFF nuclear libraries.

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.

This paper describes the contribution given by the application of the systems engineering approach to the European DEMO Breeding Blanket (BB) integration studies, focussing on the integration of Heating & Current Drive (H&CD) and Fuelling Lines (FL) systems. In particular, attention has been paid to the BB-H&CD and BB-FL interfaces identification, definition and capture of the proper interface requirements necessary to drive the integration process and, as a consequence, supply a feedback for the reciprocal design of the interconnected systems. The defined interfaces are synthetically described in the paper and the results of the interface requirements capture process are shown, discussing in detail the rationales behind the interface requirements definition. The implications of the selected interface requirements on the selection of the proposed H&CD and FL systems design options have been highlighted as well. Lastly, the open issues are discussed as well as the influence of the interface management process on the prosecution of the BB, H&CD and FL design activities.

The optimization of the aeronautical assembly lines is one of the most challenging tasks in development of a new aircraft. To date the Aeronautical companies are founding project related to the optimization of the manufacturing processes, impacting on the reduction of the non-recurring costs. The work presented focuses on methods for the implementation of the determinate assembly approach in the case of manufacturing of aeronautical structures. The methods are based on the prediction of the coupling tolerances between airframe components in order to reduce the numbers of the features to be manufactured in assembly. All that to lead to a reduction of the non-recurring costs.

The integration of the heating and current drive (HCD) systems in the EU DEMO tokamak must address a number of issues, namely space constraints in the tokamak building, remote handling requirements, breeding blanket penetration, neutron and photon radiation shielding, compliance of penetrations of the primary vacuum with safety and vacuum criteria, and a large number of loading conditions, in particular heat, electromagnetic (EM), and pressure loads in normal and off-normal conditions. A number of pre-conceptual design options for the vacuum vessel (VV) port and the port-plug are under assessment and need to be verified against all requirements and related criteria. The identification of the functional (or physics) requirements of the HCD systems remains an on-going process during the pre-conceptual design phase, hence some initial assumptions had to be made as a basis for development of the design of the VV ports and the HCD port plugs. The paper will provide an overview of present margins in the functional/physics requirements and the rationale behind the assumptions made in order to facilitate development of the pre-conceptual design options. Furthermore it will introduce the initial design concepts of the electron cyclotron (EC) Launchers and the neutral beam (NB) injectors integrated in equatorial ports. The NB duct design in DEMO and related issues such as transmission and re-ionization losses will be also addressed.

The water-cooled lithium-lead breeding blanket is in the pre-conceptual design phase. It is a candidate option for European DEMO nuclear fusion reactor. This breeding blanket concept relies on the liquid lithium-lead as breeder-multiplier, pressurized water as coolant and EUROFER as structural material. Current design is based on DEMO 2017 specifications. Two separate water systems are in charge of cooling the first wall and the breeding zone: thermo-dynamic cycle is 295–328 °C at 15.5 MPa. The breeder enters and exits from the breeding zone at 330 °C. Cornerstones of the design are the single module segment approach and the water manifold between the breeding blanket box and the back supporting structure. This plate with a thickness of 100 mm supports the breeding blanket and is attached to the vacuum vessel. It is in charge to withstand the loads due to normal operation and selected postulated initiating events. Rationale and progresses of the design are presented and substantiated by engineering evaluations and analyses. Water and lithium lead manifolds are designed and integrated with the two consistent primary heat transport systems, based on a reliable pressurized water reactor operating experience, and six lithium lead systems. Open issues, areas of research and development needs are finally pointed out. © 2019 Elsevier B.V.

The work behind this paper took place in the Eurofusion remote maintenance system project (WPRM) for the EU Demonstration Fusion Power Reactor (DEMO). Following ITER, the aim of DEMO is to demonstrate the capability of generating several hundreds of MW of net electricity by 2050. The main objective of this paper was the study of the most efficient design of the maintenance port for replacing the divertor cassettes in a Remote Handling (RH) point of view. In DEMO overall design, one important consideration is the availability and short down time operations. The inclination of the divertor port has a very important impact on all the RH tasks such as the design of the divertor mover, the divertor locking systems and the end effectors. The current reference scenario of the EU DEMO foresees a 45° inclined port for the remote maintenance (RM) of the divertor in the lower part of the reactor. Nevertheless, in the optic of the systems engineering (SE) approach, in early concept design phase, all possible configurations shall be taken into account. Even the solutions which seem not feasible at all need to be investigated, because they could lead to new and innovative engineering proposals. The different solutions were compared using an approach based on the Analytic Hierarchy Process (AHP). The technique is a multi-criteria decision making approach in which the factors that are important in making a decision are arranged in a hierarchic structure. The results of these studies show how the application of the AHP improved and focused the selection on the concept which is closer to the requirements arose from technical meetings with the experts of the RH field. © 2016 Elsevier B.V.

The computerization of manufacturing is one of the major challenges of the so-called fourth industrial revolution or Industry 4.0. Virtualization of the smart factory should provide real-time vision, control and monitoring of production through interactive dashboards and synchronization of data coming from different factory functions. The latter characteristics are particularly difficult to implement when the manufacturing core relies on traditional manual labour rather than on automation, as in the case of manual assembly. Monitoring or even controlling the manual work in real-time is extremely difficult to put into practice. Therefore, realizing the principles of Industry 4.0 in manual or semi-automatic labour contexts means developing new production control systems that involve the worker in the monitoring process without negatively affecting the production times or the psychological status of the workers. In particular, the authors propose a computer-aided production control framework based upon multimedia manuals and smart completeness control systems that can be used to implement the principles of Industry 4.0 in manual or semi-automatic work environments. This technology has been successfully tested in laboratory on the basis of a real industrial case study. The response of the testers has been positive and the outcomes in terms of increased product quality are promising. © 2018, Springer-Verlag France SAS, part of Springer Nature.

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.

The Iterative and Participative Axiomatic Design Process (IPADeP) deals with the early conceptual design stage of complex mechanical assemblies. It provides a systematic approach based on the theory of Axiomatic Product Development Lifecycle and aims to minimize the risks related to the uncertainty and incompleteness of the requirements, considering that the requirements will be refined and completed during the process. IPADeP has an iterative nature and is focused on the experience of the people involved in the design process. The functional requirements and the design parameters are conceived through brainstorming sessions and the concept selection is performed involving several experts through a Multi Criteria Decision Making technique. IPADeP has been adopted as methodology to address the early conceptual design stage of a subsystem of the DEMOnstration fusion power plant: the divertor cassette-to-vacuum vessel locking system. A first iteration was performed, resulting in the selection of a "high level" rough solution. According with IPADeP this paper presents an improvement of this solution, performing a new iteration of the process, since the system is ripe to proceed with the decomposition and zigzagging to the second level and new requirements are coming in from the development of the interfaced systems. © 2015 The Authors. Published by Elsevier B.V.

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.

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.

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.

The remote handling (RH) maintenance of components of International Fusion Materials Irradiation Facility (IFMIF) is one of the most challenging activities to be performed to guarantee the required high level of IFMIF plant availability. Among these components, the maintenance of the target assembly (TA) system appears to be critical, because it is located in the most severe region of neutron irradiation. The present European TA design is based on the so-called replaceable backplate (BP) bayonet concept. It was developed with the objective to reduce the waste material and to simplify the procedures for the target and BP replacement, thus reducing the intervention time for their substitution. The RH maintenance activity for the TA comprises a number of in situ refurbishment tasks, such as the removal of the BP, cleaning of surfaces from lithium solid deposition, inspection of the target body, installation of a new BP, and testing of the assembled system. However, there is also the possibility to replace the entire TA and to perform these refurbishment tasks offline in a dedicated hot cell. To accomplish all the refurbishment operations for the TA within the expected time for maintenance, the annual preventive maintenance period for IFMIF has been fixed in 20 days; several 3-D kinematic simulations in virtual reality environment and experimental activities aimed at developing and validating the implemented maintenance procedures for this component were carried out, in collaboration with the IDEAinVR Laboratory of CREATE/University of Naples Federico II, at the research center at ENEA Brasimone, Italy. The in situ refurbishment processes and the target replacement were simulated and tested and the feasibility of each maintenance operation was proved. In this paper, a description of the simulations and the validation activities carried out together with the main outcomes obtained are given.

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.

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.

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.

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.

This paper deals with the development of a knowledge-based engineering (KBE) approach able to support railway manufacturers in their assessments on the convenience of participating in competitive tendering and, subsequently, in the offer definition and in the designing phase. The proposed approach is based on a Decision Support System (DSS) that allows an analysis, called Adopt-Adapt-Innovate (AAI), to be made, which helps the company in the search of its products that best suit the requirements of new bids. Digital pattern techniques, configuration design methods and parametric modeling are the tools proposed to optimize the process that starts with the tender notice, passes through the offer definition and ends with the design. The paper describes the railway market logics, the proposed methodology and the first obtained results. © 2014 WIT Press.

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 focuses on a structured methodology that uses virtual reality (VR) and digital human modeling (DHM) to study maintenance procedures of industrial products. VR technologies help to highlight the most critical aspects of maintenance operations, while DHM tools allow detailing working sequences. Data coming from these analyses are then used to draw up a multimedia maintenance manual based on digital video animations, audio comments, explanatory images and written recommendations. Information is available to maintenance personnel directly on the working site through portable electronic devices. Further, web-based multimedia manuals can be updated on-line and help to shorten learning time and maintenance downtimes.

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 water-cooled lithium-lead (PbLi) breeding blanket is one of the candidate systems considered for the implementation in the European Demonstration Power Plant (DEMO) nuclear fusion reactor. This concept employs PbLi liquid metal as tritium breeder and neutron multiplier, water pressurized at 15.5 MPa as the coolant, and EUROFER as the structural material. The current design is based on the single module segment approach and follows the requirements of the DEMO-2015 baseline design. The module is constituted by a basic toroidal-radial cell that is recursively repeated along the poloidal direction where the liquid metal flows along a radial-poloidal path. The heat generated by the fusion reactions is extracted by means of separate cooling systems for the breeding zone and the first wall. A back supporting structure is dedicated to withstand loads of the module during normal and off-normal operations. Water and PbLi manifolds are integrated with primary heat transport and tritium extraction systems. The status of the conceptual design is presented, critically discussing its rationale and main features as supported by neutronic, thermal-hydraulic, magneto-hydrodynamic, and thermo-mechanic analyses. Recent results are outlined, pointing out open issues and development needs.