The qualification of in-vessel components for a fusion power plant requires a test environment with a high flux of 14 MeV neutrons over a sufficiently large surface and volume. Performance testing and qualification of the complex design and technologies of fusion nuclear components is needed, in particular that of the tritium breeding blanket (BB). Testing in relevant conditions over a relevant time will also allow gaining the necessary confidence regarding the build-up and control of tritium inventories inside the BB, which will be an important radioactive source. An option of such a volumetric neutron source (VNS) is a beam-driven tokamak. A feasibility study of the main machine components and associated plant systems is described in this article. The machine has a major radius of 2.53 m, a single-null divertor configuration, and four tangential 120 keV beamlines that generate a fusion power of approximately 30 MW and provide current drive for a steady-state plasma scenario. The plasma is small with a minor radius of a = 0.55 m to maximize the neutron wall load, up to 0.5 MW/m(2), similar to what is targeted in ITER. Approximately 25 m(2) are available for blanket testing including 4 port plugs, which offer flexibility regarding the test module operating conditions and the implementation of instrumentation. Given the small plasma, much of the tokamak's volume is made up by the neutron shielding structures that are similarly sized as in ITER. To reduce the construction risk, ITER-like concepts were adopted for many components. In some cases, however, lessons learned from ITER led to the development of customized or innovative concepts. Due to the modest fusion power the plasma will burn <1 kg of tritium per year, which can be provided from external sources.

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.

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.

Future fusion reactors using deuterium-tritium fuel will exhibit high fluences of high-energy neutrons inside and around the reactor vacuum vessel (VV). As well as causing material damage, fusion neutrons will activate materials, the decay of which leads to radiation fields in and around the reactor after shutdown. Gamma -ray emission from activated materials is a particular radiological hazard during periods of reactor shutdown. This must be accounted for in the design of the reactor shielding to ensure that risks are reduced as low as reasonably achievable. Recent neutronics work has evaluated the shutdown dose rates (SDDRs) in the EU DEMOnstration power plant (DEMO) around the ports and throughout the cryostat, incorporating prospective shielding improvements to the VV and ports. Prior to the proposed shielding design improvements, calculations for the model including the helium-cooled pebble bed (HCPB) blanket showed that radiation leakage through the blanket and VV leads to biological-equivalent SDDRs (following 12 days' decay) above 103 & mu;Sv/h throughout the cryostat, ignoring additional contribution from radiation streaming through the port openings. Inclusion of the proposed VV changes reduces this dose rate to below 100 & mu;Sv/h. The work finds an approximate order-of-magnitude reduction in SDDR throughout the cryostat when all proposed shielding improvements are applied, leading to dose rates in the cryostat in the range of hundreds to thousands of & mu;Sv/h for the full model. The work shows that to further reduce dose rates inside the cryostat, improving the shielding performance of the ports is required, with particular emphasis on the lower port and the equatorial electron-cyclotron launcher which currently dominate the dose rates.

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.

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.

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.

EU-DEMO will be the next step in Europe after ITER on the path toward a fusion power plant. The matter injection systems have to provide the requested material in order to establish, maintain, and terminate the burning plasma. Their main function is to fuel the plasma, but other tasks are addressed as well like delivering matter for generating sufficient core radiation and divertor buffering. In the preconceptual design phase performed from 2014 to 2020, the matter injection systems, in particular pellet injection and gas injection, have been assessed. This work describes the main findings and state of the art of the matter injection systems at the transition from the preconceptual design phase to the conceptual design phase.

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.

A design integration study (KDII4) was conducted in the DEMO Pre-Concept Design Phase with the primary goal to develop a pre-concept feasibility design and concept of operation for the Vertical Blanket Segment Architecture. The primary goal was to develop two workable variants for removal of large in-vessel components. This, prompted by two earlier studies, highlighting the integration challenge finding a self-consistent DEMO design point (KDII) and secondly the complexity of operation required to ensure compatibly with the proposed port-based maintenance schemes. Initially, only a single null (SN) divertor configuration was considered, but due to additional identified issues and technical challenges centred on the extraction of the BB segments, a double null (DN) alternative variant has also been investigated. This was prompted by other KDII's studies, especially KDII1 (Design, performance and feasibility of wall protection limiters during plasma transients) and KDII3 (Advanced Magnetic Configurations). Alternative forms of vertical maintenance architecture have been investigated by breaking down the study into Ports, In-vessel Components, Operations and Safety. SN and DN configurations were studied with split and full blankets numbering seven variants in total. Each variant was studied in some detail, comparatively assessing each on its merits. However, no viable solution has been found for down-selection. Nevertheless, the question of vertical maintenance is now better understood. This paper will describe the findings of KDII4 in conjunction with the Remote Maintenance (RM) technology work package, which suggests that the current ‘reference’ design envelope available for RM is too constrained.

Because of the complexity of the design of a Fusion Power Plant like the EU DEMOnstration power plant the study of the plant layout must proceed in parallel with the design of the major systems. This is necessary to ensure that the design of the plant incorporates from the very beginning sound considerations on: Safety and licensing, Personnel security in context with operation and maintenance, Adequate plant availability, and Safe performance operation including delivery of few hundred MW net electric power to the electrical grid. Though it is a process of trial and change that follows the design evolution, this approach allows a better and continuous control of the numerous physical and functional interfaces among the systems and structures assuring an optimization of the overall design focused on the above targets. Some of the fluid systems inside the tokamak building are activated e.g., water coolant circuits of in-vessel components containing activated corrosion products (ACPs), radioactive N-16 and N-17 isotopes generated from neutron irradiation of oxygen, as well as the Lithium-Lead (LiPb) system containing ACPs and radioactive isotopes generated by neutron activation of the LiPb alloy. The layout of the corresponding circuits must consider constraints such as shielding, inspection, maintainability and irradiation lifetime of some equipment of other plant systems located in the tokamak building, e.g. electric and electronic equipment, organic seals of valves, and relevant actuators. Furthermore the personnel exposure during operation and maintenance has to be as low as reasonable achievable. The experiences of Nuclear Power Plants and ITER (International Thermonuclear Experimental Reactor) are an important input to the layout of the DEMO coolant circuit. This paper presents some initial considerations on the criteria to be used for the layout criteria inside the tokamak building.

•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.

Currently, for the EU DEMO, two Breeding Blankets (BBs) have been selected as potential candidates for the integration in the reactor. They are the Water Cooled Lithium Lead and the Helium Cooled Pebble Bed BB concepts. The two BB variants together with the associated ancillary systems drive the design of the overall plant. Therefore, a holistic investigation of integration issues derived by the BB and the installation of its ancillary systems has been performed. The issues related to the water activation due to the 16N and 17N isotopes and the impact on the primary heat transfer systems have been investigated providing guidelines and dedicated solution for the integration of safety devices as isolation valves. The tritium retention and the permeation rates through the blanket and its ancillary systems have been also assessed taking into account different operating points both for the BB and ancillaries and comparing, when possible, the releases with the operating and safety limits. Moreover, the issues related to the tritium start-up inventory as well as the uncertainties on the Tritium Breeding Ratio (TBR) due to the integration of the auxiliary systems within the Vacuum Vessel have been also studied. Finally, the impact of the BB concepts on the safety systems like the Vacuum Vessel Pressure Suppression System is described with a particular focus on the different measures that should be implemented according to the considered concept. All these aspects are then taken into account to drive future developments during the Concept Design Phase.

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.

•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.

In the European DEMO program, the design development of the demonstration power plant (DEMO) is currently in its pre-conceptual phase. This work includes also the design development of the vacuum vessel, where lower ports are important appendices that include the primary vacuum pumping systems as major components of the vacuum pumping and fuel processing systems. This paper develops a preliminary design configuration of the primary pumping systems. It starts with a definition of requirements on which basis the design of this so-called 'pump cask design' is evolved. Furthermore, this paper states clearly what work is required in order to come up with a sound pre-conceptual design for DEMO in 2020 and a conceptual design in 2027.

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.