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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.

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

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

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

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

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

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.

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.

Acid-base and metal ions adsorptive properties of colloidal carbon-based nanostructured materials, bearing a variable number of oxygen-containing functional groups at the surface, were investigated and modeled. Hydrophilic carbonaceous nanoparticles (HNPs) with uniform dimensions and a good grade of dispersion in water, were obtained through the oxidation of a carbon black with nitric acid, at a temperature of 100 °C, for different reaction times (2–24 h). The acid-base properties of the functional groups were evaluated in water suspension by potentiometry and infrared spectroscopy. By means of X-ray fluorescence, infrared spectroscopy and differential pulse voltammetry the interaction properties of HNPs with Cd2+, Zn2+, Pb2+ and selected Ln3+ ions have also been investigated. The experimental data, collected in the 2.0–7.0 pH range, suggest that the adsorption of metal ions on HNPs is controlled by chemical adsorption involving the strong complexation of metal ions with the carboxylic groups on the surface of HNPs. Particularly, the high complexation of Ln3+ ions opens the exploitation of HNPs in solid phase extraction for preconcentration, separation and determination of rare earth elements and actinides.

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).