Abstract: Both EU-DEMO and CFETR need a workable replacement scheme for their tritium breeding blanket (BB). The radioactive environment dictates that all associated operations must be carried out through remotely controlled tools. Accessing and extracting the large BB segments requires their feeding pipes to be removed first from the upper vessel port and later re-installed. In the current design, each of the 80 BB segments is connected with four feeding pipes, two for cooling and two for tritium extraction. Thus, in the process of BB maintenance operation, 320 pipes must be cut, removed and rewelded. The developed concept for this task is presented here. It aims at reducing the associated plant downtime, thereby increasing the overall plant availability. Features underpinning this integrated BB pipe service concept are: (i) parallel pipe service operations, (ii) pipes in each upper port are grouped in a pipe forest and handled as a single component, (iii) the configuration of the individual pipes is standardized such that cutting and joining locations are aligned and with good accessibility from the top, (iv) the number of pipe sizes is limited to two, reducing the number of required tool sets, and (v) the same pipe configuration is adopted in each of the 16 upper ports. The paper will present design solutions and the progress on the manufacturing of prototypes developed for the challenging cutting and welding tasks from both within and outside, as well as the leak detection methodology and pipe stub handling. Prototypes will be used to perform design validation and verification on dedicated test benches currently being implemented in close collaboration between European laboratories and the Comprehensive Research fAcility for Fusion Technology (CRAFT) at ASIPP in China.

Keywords: Breeding blanket | CFETR | DEMO | Feeding pipes | Remote maintenance

Abstract: Human-Robot Interaction (HRI) is an interdisciplinary research field aiming to study and develop platform in which robots (and in particular collaborative robots, also called cobots) can interact and collaborate with humans to execute dedicated tasks. Usually, cobots are intended as passive robotic devices for direct collaboration with a human operator within a shared workspace. They are designed to be used in open and uncontrolled environments; the robot shall be able to adapt its behaviors to the dynamic input of the surrounding environment. In this optic, Mixed Reality (MR) can play a crucial role supporting the flow of data between the actors (cobot and human) working in the shared environment, it can offer a simply, but remarkably advanced, communication interface between human and robot. Thanks to MR, tools to allow human operators, without particular experience or knowledge of robotics, to easily interact with the cobot can be developed. Our work is focused on development of a MR platform that integrates cutting-edge technologies, i.e. a Head Mounted Display (HMD), and a cobot in a shared environment. The experimental setup includes the Microsoft’s Mixed Reality HMD HoloLens 2 and the Franka Emika Robot System.

Keywords: Cobot | Head Mounted Display | Human Robot Interaction | Mixed Reality | Robotics

Abstract: The DEMO breeding blanket (BB) must be replaced during the machine lifetime due to the material degradation caused by the neutron irradiation. The large BB segments can therefore be removed through the upper ports of the vacuum vessel by a remotely operated transporter. The size of these ports is however restricted by the magnetic coils causing some of the BB segments to be accessible only on their extremities. The lifting point of these BB segments therefore is away from their centers of gravity also requiring the transfer of bending moments. A concept of the BB transporter was developed recently [1]. It has the required payload capacity and is capable of carrying out also the tilting maneuvers required to extract the BB segments from the VV. The gripper interlock is the interface to the BB segments and is described in this article including the function of its locking mechanism. It has the tightest space constraints of all BB transporter components, and its design is particularly challenging given the large loads to be transferred. The basic concept of the gripper interlock resembles a massive pin with a diameter of approximately 500 mm that is inserted into a countersunk hole in the backside of the BB segment and then locked by an actuated mechanism. The concept allows on the one hand the transfer of large bending moments. The engagement is on the other hand more challenging as compared to the hook of a conventional crane that is required to transfer vertical loads only. In addition the gripper interlock must be designed according to the rules defined for lifting equipment in nuclear power plants and considering increased requirements regarding qualification and in-service inspection since its failure can cause a load drop with the potential to damage the primary confinement.

Keywords: DEMO | Remote handling | Remote maintenance | Tokamak

Abstract: The experimental evaluation of welding-induced distortion is a topic of great interest to both the scientific and industrial communities. With the aim of addressing a specific need in an industrial context, this paper investigates the capabilities of a stereo–Digital Image Correlation (stereo–DIC) system to measure the weld-induced distortion of the front-plate of a bogie train bolster subassembly. Currently, the deviation from planarity of this surface is measured at less than five points using a CMM in the post-weld cooled state. An additional machining process is then used to bring the surface within the tolerance required to join the welded assembly to the train body through a threaded flange. The paper shows that DIC can provide accurate full-field distortion and strain maps over the entire 588 mm × 308 mm surface of the front plate. The distinct strength of DIC over the currently used inspection technique is its ability to provide highly spatially dense data that are unaffected by rigid body motion. This can be advantageous in terms of saving time in the post-weld inspection and reducing errors in the finishing process. In addition, DIC capabilities revealed important information that was not available from the CMM, such as the full-surface map of the initial deviation of the plate from its nominal geometry and its asymmetric deformation after welding. Finally, the full-field nature of the experimental data obtained allows for seamless integration with FE thermo-mechanical simulations for numerical model validation, stress calculation, and optimization of geometry and technological processes.

Keywords: digital image correlation | full-field distortion and strain map | traction frame | welding-induced distortion

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

Keywords: breeding blanket | DEMO | integration | WCLL

Abstract: The breeding blanket (BB) segments are by far the largest in-vessel components of DEMO. For their remote replacement through the upper vertical ports of the vacuum vessel (VV) recently a new concept has been developed, [1]. The concept minimizes the spread of contamination as all in-vessel operations are carried out from within a cask that is sealed to the VV and located within a sealed room providing a second confinement barrier inside the nuclear building. The removal of the BB segments from the VV is carried out by a BB transporter that is operated on the elevator system of the >20m higher cask. The limited available space makes the compact design solutions that have been developed critical to the overall concept. The BB transporter is designed according to nuclear design codes and for high payloads since the BB segments may weigh up to 180 tons. Due to the eccentric engagement points on the backside of the BB segments and due to seismic accelerations, that need to be considered, too, the BB transporter resists also to bending moments. It can carry out translational as well as tilting movements as required to disengage the BB segments from their supports and to remove them through the upper VV port. The main requirements regarding integration, BB manipulation and structural integrity have been verified. Next development steps need to include further design improvements, integration of in-vessel position survey, definition and control of motion actuations, supply cable routing, the development of rescue and recovery scenarios as well as the validation in relevant test facilities. This article describes the design of the BB lifting tools including several modifications following a set of analyses that were recently performed.

Keywords: DEMO | remote handling | remote maintenance | tokamak

Abstract: The maintenance of the DEMO Breeding Blanket (BB) remotely is a crucial aspect in development of the DEMO power plant. It is a challenge due to the huge mass of the BB segment of about 180 tons. A new concept for the BB transporter has recently been developed. To properly grip and manipulate each BB segment, the BB transporter has been equipped with a gripper interlock. Due to the geometry of the BB and the vacuum vessel, the attachment point on the BB segment is not aligned with its center of gravity. Hence in addition to the vertical lifting load, large moments about the horizontal axes need to be reacted. The work discussed here concerns the structural analysis conducted on the gripper interlock; its structural integrity has been checked against the most severe load conditions that include also seismic loads according to the EN13001. Elastic analyses were performed using a finite element model in accordance with EN 13001-3-1:2012 + A2:2018, Cranes - General Design - Part 3–1: Limit States and proof competence of steel structure. The effect of the gap sizes at the contact surfaces between gripper interlock and BB after engagement as well as the effect of different friction coefficients on the sliding areas were assessed. The improvements of the design based on the structural analysis are presented, too.

Keywords: Breeding blanket transporter | DEMO | Gripper interlock | Remote maintenance

Abstract: This article describes the design and configuration of the DEMO Breeding Blanket (BB) feeding pipes inside the upper port. As large BB segments require periodic replacement via the upper vertical ports, the space inside the upper port needs to be maximized. At the same time, the size of the upper port is constrained 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. The BB feeding pipes inside the vertical port need to be removed prior to BB maintenance, as they obstruct the removal kinematics. Since they are connected to the BB segments on the top and far from their vertical support on the bottom, the pipes need to be sufficiently flexible to allow for the thermal expansion of the BB segments and the pipes themselves. The optimization and verification of these BB pipes inside the upper port design are critical aspects in the development of DEMO. This article presents the chosen pipe configuration for both BB concepts considered for DEMO (helium- and water-cooled) and their structural verification for some of the most relevant thermal conditions. A 3D model of the pipes forest, both for the Helium-Cooled Pebble Bed (HCPB) and Water-Cooled Lithium Lead (WCLL) concepts, has been developed and integrated inside the DEMO Upper Port (UP), Upper Port Ring Channel, and Upper Port Annex (UPA). A preliminary structural analysis of the pipeline was carried out to check the structural integrity of the pipes, their flexibility against the thermal load, their internal pressure, and the deflection induced by the thermal expansion of the BB segments. The results showed that the secondary stress on the hot leg of the HCPB pipeline was above the limit, suggesting future improvements in its shape to increase the flexibility. Moreover, the WCLL concept did not have a critical point in terms of the secondary stress on the pipeline, since the thicknesses and the diameters of these pipes were smaller than the HCPB ones.

Keywords: Breeding Blanket | CAD | DEMO | FEM | Upper Port

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

Keywords: DEMO | GLC | ITER | PAV | TER | WCLL BB

Abstract: The development of a remote maintenance concept to replace DEMO in-vessel components after completion of their lifecycle or in case of failure is fundamental to the successful implementation of the EU fusion roadmap. The replacement of the hot breeding blanket (BB), by far the largest in-vessel component, at the end of its lifecycle is particularly important. This includes the removal from the reactor, the transport to the active maintenance facility (AMF) where the BB is decontaminated and prepared for storage as radioactive waste and the preparation and installation of the new BB. Significant effort is made to control and minimize the spread of contamination. All operations are therefore carried out in sealed rooms and corridors. The high mass of the BB segments requires all remote handling equipment to be capable of handling high payloads of more than 100 tons. It must also operate within tight space and based on impaired feedback from control sensors in the radioactive environment. At the same time, it must be highly reliable in accordance with nuclear requirements and be recoverable in case of failure. Some concepts of BB lifting devices were investigated in the past [1] (Keep et al., 2017), but were discontinued due to insufficient payload capacity. Thus, the vertical maintenance of the BB was identified as one of DEMO's key design integration issues since failure to develop a feasible concept would potentially require major changes to the tokamak architecture [2] (Bachmann et al., 2020). A new study had been initiated with a focus on structural integrity and efficient load transfer from the BB through the RH equipment to the VV upper port. A concept of the BB transfer cask and the BB transporter resulting from this study is presented in this article together with a conceptual study of the layout of the tokamak building and the AMF. Studies of alternative concepts for in-vessel maintenance are conducted in parallel but will not be described here.

Keywords: Demo | Remote handling | Remote maintenance | Tokamak

Abstract: 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.

Keywords: breeding blanket | FEM analysis | HELIAS | Stellarator | structural design | thermo-mechanics

Abstract: 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.

Keywords: CAD | Design methodologies | Industry 4.0 | Internet of things (IoT) | Systems engineering | Welding process

Abstract: 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 major international experiment and it involves several research institutes from several countries. In such a project, a Systems Engineering (SE) approach is requested 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 requirements 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 requirements definition and assessment from stakeholders needs. The procedure is based on a demand/response framework and it is developed to obtain ITER requirements. 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.

Keywords: ITER tokamak | Requirements engineering | SMART requirements | Systems engineering

Abstract: This article describes the DEMO cryostat, the vacuum vessel, and the tokamak building as well as the system configurations to integrate the main in-vessel components and auxiliary systems developed during the Pre-Conceptual Design Phase. The vacuum vessel is the primary component for radiation shielding and containment of tritium and other radioactive material. Various systems required to operate the plasma are integrated in its ports. The vessel together with the external magnetic coils is located inside the even larger cryostat that has the primary function to provide a vacuum to enable the operation of the superconducting coils in cryogenic condition. The cryostat is surrounded by a thick concrete structure: the bioshield. It protects the external areas from neutron and gamma radiation emitted from the tokamak. The tokamak building layout is aligned with the VV ports implementing floors and separate rooms, so-called port cells, that can be sealed to provide a secondary confinement when a port is opened during in-vessel maintenance. The ports of the torus-shaped VV have to allow for the replacement of in-vessel components but also incorporate plasma limiters and auxiliary heating and diagnostic systems. The divertor is replaced through horizontal ports at the lower level, the breeding blanket (BB) through upper vertical ports. The pipe work of these in-vessel components is also routed through these ports. To facilitate the vertical replacement of the BB, it is divided into large vertical segments. Their mechanical support during operation relies on vertically clamping them inside the vacuum vessel by a combination of obstructed thermal expansion and radial pre-compression due to the ferromagnetic force acting on the breeding blanket structural material in the toroidal magnetic field.

Keywords: Breeding blanket | Cryostat | DEMO | Design integration | In-vessel components | Tokamak | Vacuum vessel

Abstract: 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.

Keywords: Breeding blanket | DEMO | Nuclear fusion | Nuclear reactor

Abstract: 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.

Keywords: Breeding blanket | DEMO | MCNP | Neutronics | Nuclear analysis | WCLL

Abstract: 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).

Keywords: CPS | ITER | PbLi loop | TES | WCLL TBS | WCS

Abstract: The eutectic alloy Lithium Lead (LiPb) enriched at 90 % in 6Li is the breeder material for one of the candidate European Breeding Blanket (BB) concepts. Currently under investigation for DEMO reactor, the Water Cooled Lithium-Lead (WCLL), and for the WCLL Test Blanket Module (TBM) that will be qualified in the ITER reactor. The LiPb alloy is used as tritium breeder, neutron multiplier and tritium carrier. The design of the LiPb loops is currently under study and the conceptual design of the main loop components has been completed. For this reason, it becomes mandatory to proceed with the integration of the LiPb loops in the EU DEMO Tokamak building, checking the consistency of the different systems design to be integrated in DEMO reactor building. CAD design and integration of the entire LiPb loops are shown taking into account the building areas assigned, the interfaces with the other systems and the requirement related to the LiPb loop functions. An initial layout of the pipework and the position of the main components have been defined on the basis of the following design requirements: (I) gamma radiation shielding of the components and the pipework; (II) target flow velocity of the LiPb; (III) thermal expansion of the pipes; (IV) possibility to drain the entire loop; (V) redundancy of the loops; (VI) remote maintenance; (VII) position in the building and dimensions of the storage tanks. The 3D model of the entire loops has been provided and integrated in DEMO Tokamak building pointing out the issues related to the interfaces with the other systems and with the building itself.

Keywords: Breeding blanket | CAD | DEMO | Integration | Lithium lead | Piping design | Water cooled lithium lead

Abstract: The paper presents a conceptual design and its structural verification of the cryostat for the DEMOnstration Fusion Reactor (DEMO). The cryostat is a large pressure vessel providing the vacuum required to operate the superconducting coils at cryogenic temperatures. Cryostats of existing machines typically are cylindrical and self-support the external pressure. In a nuclear machine, like DEMO, a massive bioshield will enclose the reactor providing radiological protection to maintenance areas in the primary building. The proposed design makes use of the bioshield to support the cryostat which allows substantially reducing the amount of steel needed for its construction. The cryostat is a conventional pressure vessel and designed according to ASME VIII, Div. 2. Linear and nonlinear structural and thermal-structural FEM assessments show that the proposed conceptual cryostat design provides both, the required membrane strength to withstand the external pressure as well as the required flexibility to allow the thermal contraction in case of a loss of vacuum event causing the cryostat to cool down. However, the relatively thin shell is not capable of bearing any significant internal overpressure. Therefore, a rupture disk in the cryostat to release Helium into the building in case of large internal leaks of liquid Helium is required.

Keywords: DEMO cryostat | Design | Finite element method | Structural verification

Abstract: 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).

Keywords: Breeding blanket | CAD | DEMO | FEM | Ratcheting | Vacuum vessel

Abstract: 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.

Keywords: Breeding blanket | DEMO | MCNP | Neutronics | Nuclear analysis | WCLL

Abstract: The equatorial outboard limiters (also called outboard midplane limiters (OMLs)) are an essential part of the DEMO wall protection concept. Limiters are foreseen in different areas of the DEMO first wall, namely in the equatorial ports, on the high-field side, in vertical ports and additional protection limiters between equatorial and lower ports. The limiters shall prevent the plasma to touch the first wall of the breeding blankets during all plasma transients. The port integration concept of the OMLs, used for plasma ramp-up/-down, is explained including (i) thermal, structural and electromagnetic loads, (ii) neutronic requirements and related material properties, (iii) remote handling considerations, (iv) space and mass constraints and (v) the required alignment precision to allow equal distribution of the heat exposure amongst the individual of the plasma facing (PFC) limiter components. While the hot fusion plasma during ramp-up is impinging directly on the limiter, its PFC components temperature is rising and can be measured by means of either thermocouples or by infrared (IR) thermography an estimation of the heat flux on the contact point can be made. This is the basis for the proposed alignment strategy.

Keywords: DEMO | Equatorial limiter | Infrared thermography | ramp-up/-down limiter | Wall protection

Abstract: 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.

Keywords: Aeronautical assembly process | Determinant Assembly | Determinate Assembly | Hole to hole | Statistical distribution | Tolerance prediction

Abstract: 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.

Keywords: Aeronautical products | Assembly process | CAD | Determinant Assembly | Determinate Assembly | Hole to hole | Statistical distribution | Tolerance prediction

Abstract: 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.

Keywords: Axiomatic Design | Design methods | Systems Engineering | Tokamak design

Abstract: 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.

Keywords: DEMO BB integration | FL | H&CD | Interface requirements | Systems engineering

Abstract: 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.

Keywords: CAD | DEMO | Integration | Upper Port | Vacuum vessel

Abstract: 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.

Keywords: CAD | DTT | EU-DEMO | FEM | Fusion reactor | Structural analysis

Abstract: 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.

Keywords: EU-DEMO tokamak | High Field Side high-speed pellet injection | Straight guide tubes

Abstract: 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.

Keywords: Electron cyclotron | Neutral beam injection | Plasma heating and current drive | Port integration

Abstract: 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.

Keywords: Breeding blanket | DEMO | EUROfusion | WCLL

Abstract: 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.

Keywords: DEMO | FEM | Systems Engineering approach | WCLL Breeding Blanket

Abstract: 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.

Keywords: CAD | Determinant Assembly | Ergonomics | Jig-less approach | Tolerance statistical prediction | Variational assemblies

Abstract: 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.

Keywords: DEMO | Electromagnetic analysis | FEM | Upper Port | Vacuum vessel

Abstract: 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.

Keywords: Cyber-physical systems | Industry 4.0 | Interactive electronic technical manuals | Production monitoring | Smart manufacturing

Abstract: 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.

Keywords: Balance of plant | Breeding Blanket | In-vessel and ex-vessel components

Abstract: 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.

Keywords: DEMO | MCNP | Neutronics | Nuclear | Shielding | TBR | WCLL

Abstract: 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.

Keywords: Breeding blanket (BB) | Demonstration Power Plant (DEMO) | fusion reactor design | liquid metal technology

Abstract: 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.

Keywords: breeding blanket | DEMO | WCLL

Abstract: 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.

Keywords: 3D CAD modeling | Conceptual design | Cryostat | FEM | First wall | Mechanical analysis | Vacuum vessel

Abstract: 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.

Keywords: Breeding Blanket | Fuelling systems | Heating systems | In-vessel components | Vacuum Vessel

Abstract: 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.

Keywords: Breeding blanket | DEMO | WCLL

Abstract: 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.

Keywords: Conceptual design | DEMO breeding blanket | Figure of merits | Poloidal segmentation | Water cooled lithium lead

Abstract: As indicated in the European Fusion Roadmap, the main objective of the Divertor Tokamak Test facility (DTT) is to explore alternative power exhaust solutions for DEMO so as to mitigate the risk that the conventional divertor based on detached conditions to be tested on the ITER device cannot be extrapolated to a fusion reactor. The issues to be investigated by DTT include: • demonstrate a heat exhaust system capable of withstanding the large load of DEMO in case of inadequate radiated power fraction;• close the gaps in the exhaust area that cannot be addressed by present devices;• demonstrate that the possible (alternative or complementary) solutions (e.g., advanced divertor configurations or liquid metals) can be integrated in a DEMO device. In this paper, we describe a proposal for such a DTT, presented by ENEA in collaboration with a European team of scientists. The selection of the DTT parameters (a major radius of 2.15 m, an aspect ratio of about 3, an elongation of 1.6-1.8, a toroidal field of 6 T, and a flat top of about 100 s) has been made according to the following specifications: • edge conditions as close as possible to DEMO in terms of dimensionless parameters;• flexibility to test a wide set of divertor concepts and techniques;• compatibility with bulk plasma performance.• an upper bound of 500 M€ for the investment costs. This paper illustrates this DTT proposal showing how the basic machine parameters and concept have been selected so as to make a significant step toward the accomplishment of the power exhaust mission.

Keywords: Design | Divertor | Tokamak devices

Abstract: 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.

Keywords: DEMO | Elasto-plastic analysis | FEM | Vacuum vessel

Abstract: The EUROfusion Consortium established in 2014 and composed by European Fusion Laboratories, and in particular the Power Plant Physics and Technology department aims to develop a conceptual design for the Fusion DEMOnstration Power Plant, DEMO. With respect to present experimental machines and ITER, the main goals of DEMO are to produce electricity continuously for a period of about 2 h, with a net electrical power output of a few hundreds of MW, and to allow tritium self-sufficient breeding with an adequately high margin in order to guarantee its planned operational schedule, including all planned maintenance intervals. This will eliminate the need to import tritium fuel from external sources during operations. In order to achieve these goals, extensive engineering efforts as well as physics studies are required to develop a design that can ensure a high level of plant reliability and availability. In particular, interfaces between systems must be addressed at a very early phase of the project, in order to proceed consistently. In this paper we present a preliminary design and integration study, based on physics assessments for the EU DEMO1 Baseline 2015 with an aspect ratio of 3.1 and 18 toroidal field coils, for the DEMO port plugs. These aim to host systems like electron cyclotron heating launchers currently developed within the Work Package Heating and Current Drive that need an external radial access to the plasma and through in-vessel systems like the breeder blanket. A similar approach shown here could be in principle followed by other systems, e.g. other heating and current drive systems or diagnostics. The work addresses the interfaces between the port plug and the blanket considering the helium-cooled pebble bed and the water cooled lithium lead which are two of four breeding blanket concepts under investigation in Europe within the Power Plant Physics and Technology Programme: the required openings will be evaluated in terms of their impact onto the blanket segments thermo-mechanical and nuclear design considering mechanical integration aspects but also their impact on tritium breeding ratio. Since DEMO is still in a pre-conceptual phase, the same methodology is applicable to the other two blanket concepts, as well.

Keywords: DEMO | electron cyclotron heating | integration | nuclear fusion | port plug

Abstract: The main goal of the Divertor Tokamak Test facility (DTT) is to explore alternative power exhaust solutions for DEMO. The principal objective is to mitigate the risk of a difficult extrapolation to fusion reactor of the conventional divertor based on detached conditions under test on ITER. The task includes several issues, as: (i) demonstrating a heat exhaust system capable of withstanding the large load of DEMO in case of inadequate radiated power fraction; (ii) closing the gaps in the exhaust area that cannot be addressed by present devices; (iii) demonstrating how the possible implemented solutions (e.g., advanced divertor configurations or liquid metals) can be integrated in a DEMO device. In view of these goals, the basic physical DTT parameters have been selected according to the following guidelines: (i) edge conditions as close as possible to DEMO in terms of dimensionless parameters; (ii) flexibility to test a wide set of divertor concepts and techniques; (iii) compatibility with bulk plasma performance; (iv) an upper bound of 500 M€ for the investment costs.

Abstract: 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.

Keywords: IFMIF | remote handling (RH) | target assembly (TA) | virtual simulations

Abstract: The present work focused on the development of a new approach to the concept design of DEMO Divertor Cassette (DC) Remote Handling Equipment (RHE). The approach is based on three main assumptions: the DC remote handling activities and the equipment shall be simplified as much as possible; technologies well known and consolidated in the industrial context can be adopted also in the nuclear fusion field; the design of the RHE should be based on a simply supported beam approach instead of cantilever approach. In detail, during the maintenance activities the barycentre of the DC is centred with respect to DC supports. This solution could simplify the design of RHE with a consequent reduction of the design and development costs. Moreover also the DC remote handling tasks shall be simplified in order to better manage the DC maintenance processes. For this reason the DC assembly and disassembly process has been simplified dividing the main sequences in basic movements. For each movement a dedicated tool has been conceived.

Keywords: DEMO | Divertor cassette | Maintenance | Remote handling

Abstract: In parallel with the programme to optimize the operation with a conventional divertor based on detached conditions to be tested on the ITER device, a project has been launched to investigate alternative power exhaust solutions for DEMO, aimed at the definition and the design of a divertor tokamak test facility (DTT). The DTT project proposal refers to a set of parameters selected so as to have edge conditions as close as possible to DEMO, while remaining compatible with DEMO bulk plasma performance in terms of dimensionless parameters and given constraints. The paper illustrates the DTT project proposal, referring to a 6 MA plasma with a major radius of 2.15 m, an aspect ratio of about 3, an elongation of 1.6-1.8, and a toroidal field of 6 T. This selection will guarantee sufficient flexibility to test a wide set of divertor concepts and techniques to cope with large heat loads, including conventional tungsten divertors; liquid metal divertors; both conventional and advanced magnetic configurations (including single null, snow flake, quasi snow flake, X divertor, double null); internal coils for strike point sweeping and control of the width of the scrape-off layer in the divertor region; and radiation control. The Poloidal Field system is planned to provide a total flux swing of more than 35 Vs, compatible with a pulse length of more than 100 s. This is compatible with the mission of studying the power exhaust problem and is obtained using superconducting coils. Particular attention is dedicated to diagnostics and control issues, especially those relevant for plasma control in the divertor region, designed to be as compatible as possible with a DEMO-like environment. The construction is expected to last about seven years, and the selection of an Italian site would be compatible with a budget of 500 M.

Keywords: divertor | plasma facing components | tokamak

Abstract: 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.

Keywords: DEMO vacuum vessel | Elastoplastic analysis | Finite element method (FEM)

Abstract: 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.

Keywords: AHP | Concept design | DEMO | Remote handling | Systems engineering

Abstract: 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.

Keywords: CAD-FEA associativity | Conceptual design | DEMO vacuum vessel | Surface modelling

Abstract: 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.

Keywords: Breeding blanket | Integration in ITER | Tritium extraction and management

Abstract: 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.

Keywords: Digital Mock-Up | Divertor | FAST | Finite element EM and mechanical analyses | Remote handling

Abstract: 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.

Keywords: CAD | Design for maintainability | Digital human modeling | Piping layout design

Abstract: 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.

Keywords: Axiomatic Design | Conceptual Design | fusion engineering | IPADeP

Abstract: 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.

Keywords: Decision support system | Digital pattern | Knowledge-based engineering | Parametric modeling

Abstract: The paper focuses on the application of the Theory of Inventive Problem Solving (TRIZ) to divertor Remote Handling (RH) issues in Fusion Advanced Studies Torus (FAST), a satellite tokamak acting as a test bed for the study and the development of innovative technologies oriented to ITER and DEMO programs. The objective of this study consists in generating concepts or solutions able to overcome design and technical weak points in the current maintenance procedure. Two different concepts are designed with the help of a parametric CAD software, CATIA V5, using a top-down modeling approach; kinematic simulations of the remote handling system are performed using Digital Mock-Up (DMU) capabilities of the software. The evaluation of the concepts is carried out involving a group of experts in a participative design approach using virtual reality, classifying the concepts with the help of the Analytical Hierarchy Process (AHP). © 2013 Elsevier B.V.

Keywords: AHP | Concept design | FAST tokamak | Interactive design | Remote handling | TRIZ

Abstract: 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. © 2013 Springer-Verlag France.

Keywords: Design for maintainability | Digital human modelling | Multimedia maintenance manuals | Virtual design review

Abstract: 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.

Keywords: Control | DEMO | FAST | ITER | MHD modes | Toroidal field ripple