Mechanical support concept of the DEMO breeding blanket

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

Matter Injection in EU-DEMO: The Preconceptual Design

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

Containment structures and port configurations

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

Integration concept of an Electron Cyclotron System in DEMO

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

Integration Concept of an Electron Cyclotron System in DEMO

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

Design of the gripper interlock that engages with the DEMO breeding blanket during remote maintenance

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 mo-ments. 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

Conceptual study of the remote maintenance of the DEMO breeding blanket

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