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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.

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.

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€.

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

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

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 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.

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.

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.

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).

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.

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.

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.

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 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.

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 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 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.

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. © 2017 Karlsruhe Institute of Technology (KIT).