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• Description of layout of active maintenance facility. • Prediction of BB transporter deformation. • Overview of the various integration issues. 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. [ABSTRACT FROM AUTHOR]

The EU-DEMO (referred to as DEMO) tokamak complex currently consists of three buildings, like in ITER: the tokamak building, the tritium building and the diagnostics building. The tokamak building houses the tokamak itself and the numerous plant systems that interface with the necessary systems to produce and control the plasma. It is designed to permit assembly, operation and maintenance of the DEMO tokamak. The vacuum vessel is organized in 8 sectors, with radial ports at the lower and equatorial levels and one vertical upper port. The general architectural structure is arranged around the tokamak with a cylindrical bioshield of 2m thickness around the cryostat and floor levels corresponding to the cryostat penetrations. Additional levels are used for the integration of auxiliary equipment for the various plant systems and for accident mitigation systems. Currently, the tokamak building also represents the final nuclear confinement barrier for the radioactive material towards the environment and the public. This safety function requires the plant and safety systems to limit the release of radioactive substances during normal operation and in accidental conditions well below the safety limits. The complexity of a fusion power plant, like DEMO, with regards to integration of plant systems is much higher than that in a fission power plant due to the larger number of plant systems. Three main criteria drive the integration work of the plant systems inside the tokamak building: (i) safety requirements, (ii) functional requirements of the plant systems themselves and (iii) the maintenance approach. Cost considerations are also taken into account together with the normal and accidental environmental conditions in the various areas of the tokamak building that might challenge the qualifications of structures, systems and components (SSC). The layout of the tokamak building has to be further developed in the Concept Design Phase to follow the plant design evolution providing feedback to the various designers in order to assure that DEMO meets the cited design criteria with an optimized and licensable layout of the most complex nuclear building. [ABSTRACT FROM AUTHOR]

• Description of blanket integration concept. • Stress calculation of breeding blanket segments. • Prediction of Breeding blanket deformation. • Identification of custom-machining concept and prediction of achievable tolerances. • Overview of the various integration issues associated with the design of in-vessel components. The DEMO tokamak architecture is based on large vertical breeding blanket (BB) segments that are accessed from a maintenance hall above the tokamak and are vertically replaced through large upper ports of the vacuum vessel (VV). The feasibility of the BB segments mechanical supports is a prerequisite of this vertical segment architecture. Their design directly impacts on the removal kinematics and the remote handling operations required for release and engagement. The supports must withstand large forces acting on the BB in particular due to electromagnetic (EM) loads. At the same time, they must ensure a sufficiently precise positioning of the BB first wall. Their design also takes into account the significant thermal expansion of the blanket segments that are operated at high temperature avoiding excessive support reaction forces. The BB support concept described in this article does not require fasteners or electrical straps to the VV and therefore much reduces the complexity of the BB remote replacement – a valuable characteristic that would make this concept a milestone in meeting one of the goals defined for the DEMO project: to develop a maintainable fusion power plant design [1]. Each blanket segment is individually supported by the VV without any physical contact to the other blankets or in-vessel components. It relies instead on vertical pre-compression inside the VV due to obstructed thermal expansion and radial pre-compression due to the ferromagnetic force acting on the BB material in the toroidal magnetic field. The verification process did not identify show stoppers. Nonetheless, a further evolution of the concept is required including design improvements to mitigate the high stress levels found in the inboard blankets during plasma disruptions. The fact that no excessively high support reaction forces or large BB deflections were found suggests though that the further development of the concept could be successful. [ABSTRACT FROM AUTHOR]

• Description of design concept of breeding blanket gripper interlock. • Description of manufacturing and assembly. • Structural integrity verification of breeding blanket gripper interlock. • Discussion of engagement. 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. [ABSTRACT FROM AUTHOR]

Due to the limited irradiation lifetime of the structural material used for in-vessel components in DEMO, it will be necessary to replace all breeding blankets within the given planned maintenance window in order to meet DEMO availability targets (Romanelli, 2012; Ellis, 2017) [ 1 , 2 ]. It is assumed that failure of in-vessel components cannot be excluded, whilst in-situ repair is unrealistic. Hence the replacement of individual breeder blankets must be technically feasible. As such, remote maintenance replacement of the breeding blankets is a mission critical operation. The baseline concept utilises vertical segment architecture to aid in the removal of the blankets (Iglesias, 2013) [ 3 ]. This choice impacts on the tokamak and plant architecture and also affects operational maintenance strategy. Within the EUROfusion PPPT program efforts have been made to perform cross work package investigations on eight Key Integration Issues needed to show the feasibility of the DEMO pre-concept design (Bachmann, 2017) [ 4 ]. Key Design Issue 4 is an investigation into the feasibility of the Vertical Segment Architecture blanket feasibility The present work documents the approach, current progress and developments within this investigation. This includes the strategy, identified risks and proposed solutions. [ABSTRACT FROM AUTHOR]

• Description of DEMO breeding blanket transporter and cask design concept • Identification of main loads acting on the BB transporter • Structural integrity verification of BB transporter and cask • Definition of blanket segment removal kinematics • Overview of the various integration issues 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. [ABSTRACT FROM AUTHOR]

Abstract This paper describes the progress of the DEMO Design Activities in Europe and particularly the work done to address critical design integration issues that affect the machine configuration and performance, the plant concept layout and the selection of system design and technologies. Work continues to be primarily focused on the design integration of a pulsed baseline DEMO reactor concept, but a number of alternative configurations (e.g., a double-null divertor and a snowflake divertor as well as a 'flexi' DEMO that operates initially in an inductively driven pulsed regime, with the possibility to be upgraded to a long-pulse or steady-state machine with a greater reliance on auxiliary current drive, etc.) are under preliminary study, especially to evaluate their DEMO reactor relevance. Some initial considerations are given on the strategy to implement a structured design and technology down-selection, that progressively reviews and narrows options to arrive at the DEMO plant concept that addresses major system integration risks and offers the best probability to satisfy all stakeholder mission requirements. Finally, some recent technical achievements are highlighted. [ABSTRACT FROM AUTHOR]

Highlights • Description of selected architecture of the DEMO tokamak. • Description of shielding concept. • Rationale of the segmentation of the in-vessel components. • Description a and rationale of selected maintenance strategy. • Description of integrated design of the breeding blanket including an overview of technology aspects. Abstract The EU fusion roadmap defines as one of its goals the development of a Demonstration Fusion Power Reactor (DEMO) to follow ITER. This device shall be tritium self-sufficient, produce net electricity, and acts as a component test facility to demonstrate fusion power plant relevant technologies, e.g. those of the breeding blanket. This article identifies the main DEMO requirements, introduces the rationales for the selected architecture of the DEMO tokamak, and describes how the configuration of the main tokamak components has been derived. This includes (i) the DEMO shielding concept, (ii) the segmentation of the in-vessel components and their maintenance strategy, (iii) an overview of the vessel and in-vessel component technologies, and (iv) a description of the integrated design of the breeding blanket. [ABSTRACT FROM AUTHOR]

• An overview of plant and tokamak building layout criteria of DEMO reactor are outlined. • A first preliminary layout of DEMO main systems is draft. • The main issues in arrangement systems in the tokamak building are pointed out. • The experience lesson learned from ITER are taken into account. • Possible solutions for each of those issues are briefly discussed. An early attention to the layout of both plant site and its buildings is essential in a complex plant under preliminary design as DEMO in order to meet the assigned targets, namely i) the licensing requirements ii) a good availability in delivery electricity to the grid. The layout definition has to follow several criteria that become more complex and stringent for nuclear buildings, e.g. functional, maintenance, fire protection, safety, human factors, shielding, and remote handling. The criterion As Low As Reasonable Achievable, with respect to the dose to the staff, has to be applied in design, operation, maintenance and decommissioning phases. The tokamak building, where several complex systems have to converge to the torus to create and control the plasma, to take out its energy and to produce and extract tritium, provides the second and ultimate confinement barrier between the environment and the hazardous and radioactive materials present inside that might be mobilised, in case of accident, by the high energetic fluids stored in DEMO systems. The layout criteria are focused on avoiding any challenge to the safety functions: e.g. no common mode failures of the safety classified systems for all reference design basis events. Furthermore the safety classified equipment have to maintain their safety function all over the plant life in such challenging environmental conditions; an accurate layout might allow the qualification possible, making milder the environment, e.g. defining adequate shielding and areas where radiation dose, magnetic field and accidental environmental conditions are reasonable for sensible equipment. The experience of NPPs and ITER is also recognized as the basis of such design criteria for DEMO. The paper will outline the main design basis events and the layout criteria presenting some applications for the tokamak building that reflect the recent progress of the DEMO design. [ABSTRACT FROM AUTHOR]

The pre-conceptual layout for an electron cyclotron system (ECS) in DEMO is described. The present DEMO ECS considers only equatorial ports for both plasma heating and neoclassical tearing mode (NTM) control. This differs from ITER, where four launchers in upper oblique ports are dedicated to NTM control and one equatorial EC port for heating and current drive (H&CD) purposes as basic configuration. Rather than upper oblique ports, DEMO has upper vertical ports to allow the vertical removal of the large breeding blanket segments. While ITER is using front steering antennas for NTM control, in DEMO the antennas are recessed behind the breeding blanket and called mid-steering antennas, referred to the radially recessed position to the breeding blanket. In the DEMO pre-conceptual design phase two variants are studied to integrate the ECS in equatorial ports. The first option integrates waveguide bundles at four vertical levels inside EC port plugs with antennas with fixed and movable mid-steering mirrors that are powered by gyrotrons, operating at minimum two different multiples of the fundamental resonance frequency of the microwave output window. Alternatively, the second option integrates fixed antenna launchers connected to frequency step-tunable gyrotrons. The first variant is described in this paper, introducing the design and functional requirements, presenting the equatorial port allocation, the port plug design including its maintenance concept, the basic port cell layout, the transmission line system with diamond windows from the tokamak up to the RF building and the gyrotron sources. The ECS design studies are supported by neutronic and tokamak integration studies, quasi-optical and plasma physics studies, which will be summarized. Physics and technological gaps will be discussed and an outlook to future work will be given. [ABSTRACT FROM AUTHOR]

• Description of investigation into potential alternative plasma scenarios. • Description of wall protection strategy and integrated solutions. • Description of investigation of power conversion concepts to use the heat from the pulsed tokamak for electricity generation. • Description of the key technology of the metal foil pump and it role in the DEMO fuel cycle. The EU fusion roadmap defines the development of a DEMO plant that achieves a long operation time and demonstrates tritium self-sufficiency and net electricity output. The pre-conceptual design phase will be concluded with a gate review in 2020. Eight key design integration issues have been identified as critical, either because the corresponding solution found in ITER is not suitable in DEMO or because the issues are DEMO-specific and not present in ITER. All of these will affect in their resolution the design and possibly the technology of several tokamak and plant systems or even the DEMO architecture: (i) Feasibility of wall protection limiters during plasma transients, (ii) integrated design of breeding blanket and ancillary systems, (iii) power exhaust taking advantage of advanced divertor configurations, (iv) tokamak architecture based on vertical blanket segments, (v) direct or indirect power conversion concept, (vi) configuration of plant systems in the tokamak building, (vii) feasibility of hydrogen separation in the torus vacuum pump and direct internal fuel recycling, and (viii) development of a reliable plasma scenario. For each of these issues potential solutions have been identified. In some cases alternative solutions are recognized and are addressed for risk mitigation. The key design integration issues (KDIIs) will be introduced and discussed in this article including results achieved so far and future work yet to be carried out. [ABSTRACT FROM AUTHOR]