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• ITER TBM Program plans to test 4 DEMO-relevant tritium breeding blanket systems. • There are two water-cooled and two helium-cooled Test Blanket Systems (TBS). • Each TBS is formed by various sub-systems with different functions and technologies. • The paper describes the on-going associated R&D and the main testing objectives. The ITER TBM Program foresees that two ITER equatorial ports are dedicated to the simultaneous operation and testing of mock-ups of four different concepts of tritium breeding blankets (TBB) ensuring tritium breeding self-sufficiency and extraction of high-grade heat for a demonstration fusion reactor (DEMO). The paper starts with the description of the four Test Blanket Systems (TBSs) for the initial configuration and the main on-going R&D carried out by the ITER Members (IMs) in support of the TBS designs. The second part addresses the main objectives of the ITER TBM program to be obtained during the testing campaigns in ITER in order to demonstrate the feasibility of the corresponding DEMO TBBs. It also addresses the main technological data/information that are expected to be obtained during the TBSs design and manufacturing phases and relevant for the DEMO breeding blanket designs. [ABSTRACT FROM AUTHOR]

• STEAM will experimentally validate the DEMO WCLL BoP water coolant systems. • STEAM is an experimental facility equipped with air coolers operating as heat sink. • The steady-state characterization of STEAM was performed by RELAP5/Mod3.3. • Critical aspects of the air coolers operation were investigated and discussed. • Preliminary control systems were tested for some of the main STEAM components. DEMO power station aims to demonstrate the generation of hundred MWs of electrical power from fusion reactions, then transmitted from the Tokamak reactor to the grid through the Balance of Plant (BoP). The design approach for the Water Cooled Lithium Lead (WCLL) Breeding Blanket (BB) Primary Heat Transfer Systems (PHTSs) leverages nuclear industry expertise but faces challenges due to DEMO pulsed operation and low-load periods. To assess the feasibility of these components, ENEA Experimental Engineering Division at Brasimone R.C. is designing STEAM, a facility investigating water technologies applied to the DEMO BB and BoP systems and components. STEAM is mainly composed of a primary system reproducing the DEMO WCLL BB PHTS thermodynamic conditions (15.5 MPa, 328–295 °C) and a secondary two-phase (liquid/steam) loop reproducing the DEMO power conversion system conditions (6.4 MPa, 238–300 °C). Experimental validation will reproduce steady-state and transient operation under DEMO-relevant conditions, including dedicated tests on the DEMO once through steam generator mock-up. STEAM objectives and description are presented in this paper, together with the RELAP5/Mod3.3 nodalization of the facility. The latter is used to, thermal-hydraulically characterize the facility behaviour. The outcomes of the steady-state qualification supported the optimization of the system layout appraising the performances of key components under the specified operating conditions. [ABSTRACT FROM AUTHOR]

• A lot of experience has been gained at CEA in designing neutron sensors to be applied to DEMO breeding blankets. • Self-powered neutron detectors, fission chambers, optical fission chamber, dosimetry with activation foils are considered. • An effort joining theoretical/simulation developments to experimental campaigns is needed for DEMO application. Neutron measurements are routinely performed in fission reactors. A lot of experience has been gained at CEA in designing, prototyping and testing neutron sensors and associated acquisition systems in several nuclear facilities with a wide range of demanding conditions. In this paper, the applicability of this experience to neutron diagnostics in the breeding blankets of DEMO is addressed, as a complement to the already planned diagnostic strategy. The breeding blankets are a key component of a fusion reactor such as DEMO. The expected performances and operational constraints are demanding, nevertheless some of those, as the neutron flux range and high temperature, are close to what is encountered in sodium-cooled fast neutron reactors. The strengths and drawbacks of neutron sensors are presented with respect to the performances and constraints. Self-powered neutron detectors and fission chambers benefit from a vast return of experiment in fission reactors and received recent developments in the framework of studies for sodium-cooled fast neutron reactors. The innovative optical fission chamber is discussed for its electromagnetic immunity and high dynamic range demonstrated up to seven decades. Dosimetry with activation foils is also considered for its ability to provide an absolute calibration and spectral unfolding. [ABSTRACT FROM AUTHOR]

The Broader Approach (BA) activities aim to complement the ITER project and to promote the early realization of fusion energy through research, development, and tests of technologies supporting the future demonstration fusion reactor (DEMO). These activities are implemented under the BA agreement, which was signed and ratified in 2007 between Euratom and the Government of Japan. In essence, the BA activities consist of three projects: a) The Satellite Tokamak Programme Project JT-60SA, the world's largest superconducting tokamak until ITER starts, aims to support the assembly, commissioning and preliminary operation of ITER, and carry out demonstration and optimisation of steady-state operation of advanced plasma configurations for DEMO; b) The International Fusion Energy Research Centre (IFERC), which comprises three different sub-projects: the DEMO Design Research and Development Coordination Centre to coordinate design and R&D on materials and components for DEMO, the Fusion Computer Simulation Centre for the simulations of fusion plasmas, the analysis of experimental data, modelling of ITER operation, and contribution to the design of DEMO, and the Remote Experimentation Centre to allow scientists to participate remotely in fusion experiments from its control room in Japan; c) The Engineering Validation and Engineering Design Activities for the International Fusion Materials Irradiation Facility (IFMIF/EVEDA) focused on the design and validation of key components needed for the future Fusion Neutron Source facilities to characterize materials envisioned for DEMO. After 13 years of fruitful collaboration, Euratom and Japan launched the second phase of the BA activities in 2020. It is focused on exploiting and enhancing the facilities that have already been built and on working more closely than ever with ITER as the latter advances towards the first plasma. This paper will present a synthesis of the achievements already obtained thanks to the three BA projects and will address the future developments planned in 2023 and beyond. [ABSTRACT FROM AUTHOR]

The Water-Cooled Lead Lithium (WCLL) and the Helium-Cooled Pebble Bed (HCPB) Breeding Blanket (BB) concepts are the two candidates currently poised to be chosen as the driver blanket for the EU DEMO reactor. Nevertheless, new variants are emerging with the aim of overtaking the potential showstoppers arisen during the WCLL and HCPB BB pre-conceptual design phase. Then, an intense campaign of exploratory studies has been launched in EU to investigate the potential performances of such alternative concepts. Among them, the Water-Cooled Lead-Ceramic Breeder (WLCB) BB concept is one of the most promising. It has been conceived as a trade-off (i.e. an hybrid concept) between WCLL and HCPB BB concepts, trying to take and integrating the best features of both. Since in the reference WCLL geometric layout the Top Cap (TC) region was identified as particularly critical from thermal and mechanical standpoints, so requiring specific design studies, also in the WLCB conceptual design activities a dedicated campaign of analysis has been necessary to preliminarily size the TC region components. Hence, in this paper, the preliminary thermo-mechanical assessment of the TC region of the WLCB BB alternative concept is presented. First, an initial geometric layout has been set-up on the basis of the main features of the reference WCLL BB TC region. Then, thermal analysis under nominal conditions has been performed and the results have shown that modifications were necessary to ensure the compliance with the thermal requirements of the structural material. Finally, once obtained an acceptable geometric layout, mechanical analyses were carried out in order to verify the fulfilment of the RCC-MRx structural design criteria, also considering an off-normal load case. The results showed that this variant is very promising in terms of structural behaviour, whereas further design iterations are necessary to lower the temperature of the breeder zone. • The Water-Cooled Lead-Ceramic Breeder is an innovative Breeding Blanket concept for DEMO. • The Top Cap region is identified as critical from thermo-mechanical standpoints. • Design by analysis procedure looking at the Eurofer maximum allowable temperature. • Verify the fulfilment of the RCC-MRx structural design criteria. [ABSTRACT FROM AUTHOR]

A key aspect for the achievement of the fuel self-sufficiency of future fusion power plants is the management of tritium in the different stages of the reactor fuel cycle. In particular, one of the greatest challenges for the success of liquid metal breeder concepts such as the Water-Cooled Lithium-Lead (WCLL) is the extraction of tritium from the liquid LiPb eutectic alloy (15.7 at.%Li), which fulfils the main functions of tritium breeder, tritium carrier and neutron multiplier, and the processing in a form suitable for an accurate tritium accountancy. Within this frame, significant advancements have been carried out in the last years as regards the WCLL Test Blanket System (TBS) of ITER. Tritium is mainly produced inside the LiPb and extracted through the Tritium Extraction Unit (TEU). The reference technology for the TEU of the WCLL-TBS is the Gas-Liquid Contactor (GLC) in the packed-column configuration. Tritium is then concentrated in ZrCo getter beds before being sent to the accountancy system where different technologies for both static and dynamic accountancy are adopted in order to provide precise and reliable tritium measurements, which are an essential need in view of an application to the European DEMO reactor. In this paper, the present status and the design solutions foreseen for tritium management in WCLL-TBS are presented and discussed. The implementation and optimization of the conceptual design of the lithium-lead loop (LLP), the Tritium Extraction System (TES) and the Tritium Accountancy System (TAS) have been performed in order to manage tritium concentration in LiPb and in the ancillary systems. [ABSTRACT FROM AUTHOR]

• STEAM will host the mock-up of the EU-DEMO Steam Generators. • STEAM primary loop has been investigated with RELAP5/Mod3.3 system code. • DEMO normal operations have been studied including pulse-dwell transitions. • Different primary loop management strategies have been numerically simulated. • Setting the primary average temperature minimizes the loop operative fluctuations. The Water Cooled Lithium Lead Breeding Blanket (WCLL BB) is a key candidate for the driver blanket of the European DEMO reactor, progressing toward its Conceptual phase by the end of 2027. To assess different water and lithium-lead technologies for the WCLL BB and Balance of Plant (BoP) systems, the Water-thermal-HYDRAulic (W-HYDRA) experimental platform is under development at the ENEA Brasimone Research Centre. Among the facilities constituting the new W-HYDRA multipurpose infrastructure, STEAM is going to experimentally investigate the DEMO WCLL BoP thermal-hydraulics, focusing on the Steam Generator (SG) of the Primary Heat Transfer Systems (PHTS), to qualify its performances and suitability under its unconventional operation. The paper aims at supporting the thermal-hydraulic characterization of the Steam Generator mock-up during the sudden power variations typical of a pulsed fusion reactor. The analyzed selected scenario is the operational transient dwell-pulse-dwell, which determines high thermal cycling and correspondent high thermo-mechanical stresses on the primary side components. Two control logics with their relative drawbacks have been analyzed with a RELAP5/Mod3.3 1-D model, the first regulating the primary side average temperature, the second monitoring the minimum one. The comparison of the two systems highlighted that neither approach leads to hazardous conditions for the facility. However, while the average temperature controller is characterized by reduced thermal stresses on the components, the minimum temperature controller is characterized by higher thermal gradients for the primary loop. Both methodologies will be tested in the dedicated experimental campaign, aiming at yielding insights and evaluations concerning control strategies applicable to the DEMO reactor. [ABSTRACT FROM AUTHOR]

An Electron Cyclotron (EC) Heating and Current Drive (HCD) system will be installed in the EU DEMO Tokamak to provide plasma heating and stabilization by coupling up to 130 MW of mm-wave power into specific regions of the plasma. The mm-waves will be generated by gyrotrons and propagated through a quasi-optical multi-beam mirror system (MBMS), and subsequently through corrugated waveguides, to launchers located at dedicated Equatorial Ports (EPs). The present design of the launcher features two mid steering antennas (MSAs) for plasma stabilization each consisting of three-beam layouts and steering mirrors positioned at the upper and lower locations of the EP. The proposed concept for the steering mirrors operates on the pantograph principle, a mechanical assembly of parallel linkages, which translates the movement produced by pneumatic actuators into an exact angular rotation of the mirror around an offset virtual axis. This design uses flexures with specific structural compliance as mechanical joints resulting in the desired movement of the linkage system while avoiding the tribological difficulties associated with mechanisms operating under vacuum. This paper describes the present status of the DEMO EC HCD steering mirror as well as the analyses performed to validate this design. [ABSTRACT FROM AUTHOR]

The M1 mirrors are part of the in-vessel quasi-optical Electron Cyclotron beam propagation system of the DEMO nuclear fusion reactor. These four mirrors are all non-movable and with a plane reflecting surface. Based on their functionality they are divided into two groups, the Mid-Steering Antenna (MSA) and the Multi-Beam Mirror (MBM) antenna. The MSA is responsible for controlling the neoclassical tearing modes, while the MBM is available for plasma bulk heating, current drive and thermal instability control. These mirrors are designed to withstand the thermal loads generated during operation: neutron heating, ohmic losses and stray radiation caused by the mm-wave beam reflection. To fulfil the design requirements, efficient heat removal strategies based on water cooling principles minimizing pressure drop are implemented. This paper presents the pre-conceptual design of the MBM and MSA M1 mirrors, validating the structural and geometric boundary conditions, as well as the appropriate selection of materials. Using FEA tools, thermal-hydraulic and thermo-mechanical simulations were performed with the available thermal loads to optimize the cooling parameters and perform a first stress assessment on the two M1 mirrors. • Thermal-hydraulics and thermo-mechanical simulations performed on the M1 mirrors. • The cooling channels design is mainly based on parallel spirals at the beam spots. • Ohmic losses (170 GHz) and neutron heat load were considered as thermal loads. • The inlet conditions of the water cooling line are 75 °C and 3.5 MPa. • The design ensures acceptable temperature, pressure and deformation. [ABSTRACT FROM AUTHOR]

The International Fusion Materials Irradiation Facility-DEMO Oriented NEutron Source (IFMIF-DONES) is an accelerator-based neutron source that generates high-energy neutrons via stripping processes by focusing a high-energy deuteron beam on a fast-flowing liquid lithium jet. The neutrons produced are used for the irradiation of materials foreseen in DEMO, thus providing relevant data for the design and licensing of the future fusion reactor. The complexity of such a plant is managed by a central control system that guarantees the safe supervision and control of all operations. This paper summarizes the most recent developments in the design of the IFMIF-DONES plant's Central Instrumentation and Control Systems (CICS) after the completion of the preliminary design phase. In particular, the architecture of two of the main CICS systems (namely, MPS and SCS) is described in detail, focussing on the specific design choices recently proposed. For each system, the current status of design is presented, as well as the existing and future plans for their integration in a unique control framework. [ABSTRACT FROM AUTHOR]

The Beam Driven Plasma Neutralizer (BDPN) has been proposed as a more efficient alternative to the gas neutralizer for negative-ion based Neutral Beam Injection (NNBI). In this paper we model the performance of an entire NNBI beamline with a BDPN. We simultaneously consider all the relevant physics and engineering aspects, the most important being the plasma density and degree of ionization inside the BDPN as a function of its geometry and feed gas flow, the geometrical transmission of the beamline, the dependence of the neutral gas distribution in the beamline on the geometry of the beamline components and gas flows, and the species evolution of the extracted D − beam through this neutral and charged particle distribution. Furthermore, we calculate the heat loads expected on the BDPN parts and on the NBI components located downstream of it and study the effect of the magnetic cusp field across the BDPN entrance on beamline transmission. While our results constitute an optimization only under the applied boundary conditions, we find that the beamline with a BDPN increases the system's wall plug efficiency by about 13% to 0.34 from the 0.30 estimated for a gas neutralizer. [Display omitted] • The Beam Driven Plasma Neutralizer for DEMO NBI now fits a realistic particle beam. • An optimization to maximize the neutral power at the NBI exit has been carried out. • A fair comparison with a Gas Neutralizer resulted in a 13% increase in efficiency. [ABSTRACT FROM AUTHOR]

The reflectometry diagnostic proposed for DEMO aims to measure the radial edge density profile and to provide data for the feedback control of plasma position and shape. Microwave (MW) reflectometry measurements are foreseen at 16 poloidal locations (gaps) and require plasma-facing antennas and waveguides (WGs) to route the microwaves through the upper port (UP) to the diagnostic hall. The current integration proposal is based on the Diagnostics Slim Cassette (DSC) concept, a full, 25-cm thick poloidal sector dedicated to house the antennas and WGs. This paper translates the main findings from previous studies into an updated, coherent design of the DSC, while discussing integration aspects not considered before. Neutronics simulations, thermo-mechanical analyses and electromagnetic simulations are employed to evaluate the WG deformation under operating conditions and its effect on the reflectometry measurements. Although the estimated deformations are not expected to compromise the performance of the diagnostic, this work shows that the operating conditions must be considered when optimizing the shapes of the WGs to minimize losses. [ABSTRACT FROM AUTHOR]

• Work completed by the Heating and Current Drive Workpackage for DEMO during the pre-conceptual design phase (2014-2020) • The Electron Cyclotron System is the baseline for DEMO • Description of the Electron Cyclotron System • Description of the Ion Cyclotron System • Description of the Neutral Beam Heating System The European DEMO is a pulsed device with pulse length of 2 hours. The functions devoted to the heating and current drive system are: plasma breakdown, plasma ramp-up to the flat-top where fusion reactions occur, the control of the plasma during the flat-top phase, and finally the plasma ramp-down. The EU-DEMO project was in a Pre-Concept Design Phase during 2014-2020, meaning that in some cases, the design values of the device and the precise requirements from the physics point of view were not yet frozen. A total of 130 MW was considered for the all phases of the plasma: in the flat top, 30 MW is required for neoclassical tearing modes (NTM) control, 30 MW for burn control, and 70 MW for the control of thermal instability (TI), without any specific functions requested from each system, Electron Cyclotron (EC), Ion Cyclotron (IC), or Neutral Beam (NB) Injection. At the beginning of 2020, a strategic decision was taken, to consider EC as the baseline for the next phase (in 2021 and beyond). R&D on IC and NB will be risk mitigation measures. In parallel with progresses in Physics modelling, a decision point on the heating strategy will be taken by 2024. This paper describes the status of the R&D development during the period 2014-2020. It assumes that the 3 systems EC, IC and NB will be needed. For integration studies, they are assumed to be implemented at a power level of at least 50 MW. This paper describes in detail the status reached by the EC, IC and NB at the end of 2020. It will be used in the future for further development of the baseline heating method EC, and serves as starting point to further develop IC and NB in areas needed for these systems to be considered for DEMO. [ABSTRACT FROM AUTHOR]

The Pre-Concept Design (PCD) of the Balance of Plant (BoP) systems of the EU-DEMO power plant is described in this paper for both breeding blanket (BB) concepts under assessment, namely the Water Cooled Lithium Lead (WCLL) BB and the Helium Cooled Pebble Bed (HCPB) BB. Moreover, the results of a preliminary evaluation of a number of BoP variants are discussed. This paper outlines the steps of the BoP design development, highlighting the project objectives and the strategy for their achievement under the very challenging requirements which include, among others, the intermittent nature of the DEMO plasma heat source. The main achievements during the PCD Phase will be reported together with the development plan for the Concept Design (CD) Phase to reach a mature (feasible) BoP concept for DEMO. [ABSTRACT FROM AUTHOR]

Measurements of the steady-state magnetic field of ITER and DEMO reactors will be performed by Hall sensors that will be part of the magnetic diagnostics. The sensors will contribute to the measurement of the plasma current, plasma-wall clearance, and local perturbations of the magnetic flux surfaces near the wall. The paper describes new ceramic-metal Hall sensors with an antimony sensing layer. The results of temperature tests of second-generation antimony sensors based on thick printed copper technology are presented, including testing of the long-term stability of the sensors. [ABSTRACT FROM AUTHOR]

• EU-DEMO started the enquiry for an easy-to-use solution for sealing and fixing of vacuum vessel port closure plates. • Point is to involve industry at an early stage to create incentives for future participation. • First structuring of the different thematic areas could already take place. • Solutions aim to avoid bolted connections as much as possible. • The involvement of a specialist in the design and layout of metal seals as an industrial partner has already taken place. DEMO should be as close as possible to a future fusion power plant, produce electricity and have a high level of availability. In the current Concept Design phase, the design is to be further elaborated and industry is to be increasingly involved as a partner in the development. An essential component will be the vacuum vessel (VV), which has to fulfil many tasks. It serves as a first confinement barrier and houses in-vessel components (IVCs) that are necessary for operation. The VV has numerous openings - the ports. These are closed during DEMO operation by so-called port closure plates (PCPs). All handling of the PCPs and IVCs, as well as inspection throughout the operational phase, must be done through remote maintenance tools and without the presence of personnel. Within the framework of qualified partners from industry the fastening and sealing of the PCPs is studied. Many requirements are associated with the technical implementation. As the development of DEMO is still in an early phase, the present goal is not yet the development to a high degree of maturity. Instead the review of established industrial solutions regarding their suitability to DEMO as well as the identification of adapted solutions with more attractive features for the DEMO application are aimed for. In this paper, the background as well as the status of this task is explained. [ABSTRACT FROM AUTHOR]

• Review paper summarizing years of research on fusion materials. • Advanced modifications to F82H and EUROFER. • Summarised process to validate RAFM steels for operation in DEMO IVCs. • Strategic direction for EU-DEMO materials data gathering presented. This work addresses an overview of the recent progress, associated risks and development plans of structural materials for in-vessel components (IVCs) in DEMO plants. Reduced Activation Ferritic Martensitic Steels form the primary structural materials for most DEMOs IVCs and will be the focus of the paper. An overview of EU- and J-DEMO programs RAFM steels developments is presented. We highlight the present status in their development for use in DEMO plants, with a focus on the structural materials' operational, and project-oriented, requirements. The development of materials property handbooks from high-quality data is illustrated. A process to validate these steels for operation in DEMO IVCs is summarised, revealing the pragmatic procedures ongoing, and limitations of this approach due to the synergistic operational effects on IVC materials; the use of in-situ or surveillance monitoring is outlined as a method to accommodate this limitation in synergistic effects and allow confidence in future DEMO operations. The development of advanced modifications to F82H and EUROFER are highlighted, with minor modifications leading to improved low and high-temperature operational design space being open for DEMO reactors. [ABSTRACT FROM AUTHOR]

During the EU DEMO Pre-Concept Design Phase, the remote maintenance team developed maintenance strategies and systems to meet the evolving plant maintenance requirements. These were constrained by the proposed tokamak architecture and the challenging environments but considered a range of port layouts and handling system designs. The design-driving requirements were to have short maintenance durations and to demonstrate power plant relevant technologies. Work concentrated on the in-vessel maintenance systems, where the design constraints are the most challenging and the potential impact on the plant design is highest. A robust blanket handling system design was not identified during the Pre-Concept Design Phase. Novel enabling technologies were identified and, where these were critical to the maintenance strategy and not being pursued elsewhere, proof-of-principle designs were developed and tested. Technology development focused on pipe joining systems such as laser bore cutting and welding, pipe alignment, and on the control systems for handling massive blankets. Maintenance studies were also conducted on the ex-vessel plant to identify the additional transport volumes required to support the plant layout. The strategic implications of using vessel casks, and of using containment cells with cell casks, was explored. This was motivated by the costs associated with the storage of casks, one of several ex-vessel systems that can drive the overall plant layout. This paper introduces the remote maintenance system designs, describes the main developments and achievements, and presents conclusions, lessons learned and recommendations for future work. [ABSTRACT FROM AUTHOR]

DEMO will be a fusion power plant demonstrating the integration into the grid architecture of an electric utility grid. The design of the power conversion chain is of particular importance, as it must adequately account for the specifics of nuclear fusion on the generation side and ensure compatibility with the electric utility grid at all times. One of the special challenges the foreseen pulsed operation, which affects the operation of the entire heat transport chain. This requires a time-dependant analysis of different concept design approaches to ensure proof of reliable operation and efficiency to obtain nuclear licensing. Several architectures of Balance of Plant were conceived and developed during the DEMO Pre-Concept Design Phase in order to suit needs and constraints of the in-vessel systems, with particular regard to the different blanket concepts. At this early design stage, emphasis was given to the achievement of robust solutions for all essential Balance of Plant systems, which have chiefly to ensure feasible and flexible operation modes during the main DEMO operating phases – Pulse, Dwell and ramp-up/down – and to adsorb and compensate for potential fusion power fluctuations during plasma flat-top. Although some criticalities, requiring further design improvements were identified, these preliminary assessments showed that the investigated cooling system architectures have the capability to restore nominal conditions after any of the abovementioned cases and that the overall availability could meet the DEMO top-level requirements. This paper describes the results of the studies on the tokamak coolant and Power Conversion System (PCS) options and critically highlights the aspects that require further work. [ABSTRACT FROM AUTHOR]

• Application of Serpent 2 Monte Carlo code for nuclear fusion-relevant spectra and geometries. • Application of novel weight window methods in Serpent 2 to deep shielding problems • Benchmarking of Serpent 2 nuclear responses against SINBAD experimental data and MCNP. Released in 2009, the Serpent Monte Carlo code has established itself as a highly efficient and powerful simulation code for nuclear systems analysis. Originally developed for reactor physics applications, the scope of the code now extends to coupled multi-physics simulations and photon transport. The latter has allowed adoption of the code by the fusion neutronics community following developments of a coupled neutron-photon capability in 2014 and the ability to handle complex geometry types in 2016. The code is well validated for the energy regimes and geometry types one can expect in fission reactor analysis. Over the course of recent years a benchmarking effort has been undertaken for application of the code to nuclear fusion. Compared to nuclear fission, or accelerator based applications, the underlying particle interaction phenomena differ greatly at the energies expected in a fusion reactor as well as the specific responses that are of interest. In this paper, a novel weight window generation implementation in Serpent is investigated. The applicability of this method is demonstrated for the Frascati Neutron Generator (FNG) bulk blanket and shield experiment, part of the SINBAD database, and a DEMO helium cooled pebble bed (HCPB) computational model. A comparison is performed against MCNP using weight windows generated with ADVANTG. Excellent agreement is found for the specified tallies and the significant efficiency gain using weight windows generated using both methods is comparable. A robust variance reduction method implementation is fundamental to applications to fusion neutronics and as such, this work is an important step in deployment of Serpent for this type of analysis. [ABSTRACT FROM AUTHOR]

This work presents an overview of the integrated strategy developed, as part of the DEMO Key Design Integration Issue 1 (KDII1), to protect the EU-DEMO first wall (FW) from planned and unplanned plasma transients by employing discrete limiters. The present Breeding Blanket (BB) FW design, which aims at minimizing the loss of neutrons while travelling to the breeding zone, is able to withstand steady state heat fluxes up to ≈1-1.5 MW/m² [1], which is not sufficient to guarantee its integrity for most plasma-FW direct contact. This is different from ITER, which has a FW designed for peak heat loads up to 4.6 MW/m2 [2], and it does not have the DEMO BB breeding related requirement. A series of documents was compiled in the DEMO Pre-Conceptual Design Phase, in support of the KDII1. The work presented here was presented at the 2020 DEMO Gate 1 (G1) review, and collects also the comments of the panel and the relative additional studies triggered by them. The design process, presented in this paper was adopted to systematically evaluate the impact of design changes, or new physics inputs, on the FW protection strategy and integration issues. It includes compiling the list of transients, and performing the relative plasma simulations, the design of discrete limiters and the evaluation of their capability to reduce the heat flux density on the FW, and finally a preliminary analysis of the heat loads effects on the Plasma Facing Components (PFC). All these aspects, together with preliminary limiter design, where considered since the beginning, in an integrated way. [ABSTRACT FROM AUTHOR]

This paper is part of a series of publications concerning the development of the European DEMO during the Pre-Concept Design Phase (2014-2020), and also describing the strategy for the next phase. In particular, it deals with the physics basis of the plasma scenarios employed for the definition of the various DEMO baselines released so far, and the assumptions adopted where necessary. In the course of the Pre-Concept Design Phase, some of these assumptions have been progressively replaced with the results of dedicated modelling activities or code developments in general, which are summarized here. The considered baselines, obtained with the systems code PROCESS, are the DEMO 2015, 2017 and 2018, based on an ITER-like ELMy H-mode confinement regime. In addition, since it is now essential to avoid ELMs, baselines with some of the characteristics of QH-mode and I-mode have been produced in 2019. It has been concluded that the present integrated plasma scenarios are not secure enough for an engineering design, so a strong programme to improve them is required and planned. A discussion on the main knowledge gaps, as well as the strategy to be adopted in the next phases to close them, is provided. [ABSTRACT FROM AUTHOR]

In a power plant scale fusion reactor, a huge amount of thermal power produced by the fusion reaction and external heating must be exhausted through the narrow area of the divertor targets. The targets must withstand the intense bombardment of the diverted particles where high heat fluxes are generated and erosion takes place on the surface. A considerable amount of volumetric nuclear heating power must also be exhausted. To cope with such an unprecedented power exhaust challenge, a highly efficient cooling capacity is required. Furthermore, the divertor must fulfill other critical functions such as nuclear shielding and channeling (and compression) of exhaust gas for pumping. Assuring the structural integrity of the neutron-irradiated (thus embrittled) components is a crucial prerequisite for a reliable operation over the lifetime. Safety, maintainability, availability, waste and costs are another points of consideration. In late 2020, the Pre-Conceptual Design activities to develop the divertor of the European demonstration fusion reactor were officially concluded. On this occasion, the baseline design and the key technology options were identified and verified by the project team (EUROfusion Work Package Divertor) based on seven years of R&D efforts and endorsed by Gate Review Panel. In this paper, an overview of the load specifications, brief descriptions of the design and the highlights of the technology R&D work are presented together with the further work still needed. [ABSTRACT FROM AUTHOR]

• An overview about the engineering aspects of the European DEMO wall protection strategy is given. • The background of the wall protection issue and design approaches are explained. • Recent outcomes from the interim design studies and technology R&D efforts are presented. • Focus is placed on the innovative technology achieved for a novel armor material and joining. • Results from the extreme and medium heat flux tests are also presented. Off-normal macroscopic plasma instabilities which possibly occur in a fusion power plant represent the most critical operational accidents and pose a serious risk to economically viable exploitation of fusion power. During such an instability, a huge amount of plasma energy is rapidly deposited onto the plasma-facing first wall in form of extreme heat flux. Such heavy thermal shocks can cause severe damages or a fatal failure of the first wall even by a single event. This issue is a serious design concern raising the requirement of a wall protection strategy. To cope with this issue, the European DEMO project adopted the limiter concept. The limiters equipped with tungsten sacrificial armor shall be installed at those wall areas where plasma contact is expected to be probable. To implement the limiter concept, comprehensive engineering studies have been conducted including conceptual design of the plasma facing component and technology R&D for armor material and joining. In this overview paper, the background of the wall protection issue and its implication on design and materials are explained and the outcomes from the recent project period (2019-20) are reported. The focus is placed on the technology achievement related to the development of novel lattice-type tungsten armor material fabricated by an innovative additive manufacturing process. Interim results from the extreme and medium heat flux tests and the joining trials for tungsten-to-heat sink material (copper or steel) are also addressed. The design rationale and the interpretation of the test data are supported by computational simulations. [ABSTRACT FROM AUTHOR]

In the DEMO Pre-Concept Design Phase, a novel fuel cycle architecture has been developed that aims at reducing the tritium inventory of the plant. In contrast to previous designs, the central role thereby is taken by an additional direct internal recycling loop. By already separating a significant proportion of the fuel at the pumping duct and providing this purified gas almost directly to the injection systems, the load on the tritium plant as well as the overall plant tritium inventory is reduced substantially. This paper summarizes the activities conducted in regard of integration and presents the results (aka the key design integration issues) for the design and feasibility of a pumping concept based on tritium direct recycling. The focus is on the aspects of integration of such a fuel separation function with the main technologies involved. In the functional position foreseen, these environment conditions mainly are space constraints, strong magnetic field, and the limited amount and restrictions in occupation of suitable ports. Two technical implementations for providing the fuel separation function have been identified: A metal foil pump relying on the physical phenomena of superpermeation of hydrogen through a metal membrane is able to provide continuous gas transfer. A multi-stage cryopump utilizing gas specific cryogenic trapping on several stages. Relying on gas binding, it only can operate intermittently. Within this activity, these variants are specifically reviewed for their suitability also to provide criteria for a future down selection. [ABSTRACT FROM AUTHOR]

This paper provides an overview of the methodology adopted within the EUROfusion framework covering the Pre-Concept Design Phase (2014–2020). The phase activities focused on advancing the technical basis of the design of the European DEMOnstration Fusion Power Plant (DEMO). The emphasis in the latter stages was placed upon an integrated design approach, based on systems engineering, providing a clear direction for urgent technology R&D and addressing the key design integration issues by assessing the critical systems interdependencies and uncertainties of certain design assumptions. The programme has continued to benefit from the ITER return of experience, in terms of design, licensing, and construction and DEMO will maintain a strong link with the ITER schedule in the coming design phases. The growing need for a structured and transparent review process gave rise to the introduction of a phase-gate review process, the implementation of which led to the first formal DEMO-wide gate review held at the end of the Pre-Concept Design Phase. The Gate Panel's recommendations provided clarity on the key priorities for the programme moving forward and serve to underpin the planning for the next phase. [ABSTRACT FROM AUTHOR]

In the preparation for the Concept Design Phase and a new European Framework Programme starting in 2021, EUROfusion developed a DEMO project charter, consisting of updates to the project governance and the high level objectives, schedule and scope of work. This proposal, together with the more detailed proposals covering the scope of system design and technology R&D across the breadth of the DEMO project, forms the plan forward for DEMO and was submitted as input for approval to the first DEMO gate review (G1). Following the successful completion of the gate and the associated recommendations provided, the plan forward was modified accordingly. This paper details the principle elements of the implementation and organisation of the DEMO design and R&D activities in the Concept Design Phase. [ABSTRACT FROM AUTHOR]

• DEMO NBI requires an acceleration voltage up to 1 MV, as for the ITER NBI. • ITER AGPS system based on multiple conversion stages with gas-insulated parts. • Study of MMC topology, adopted in HVDC systems, and its application to AGPS systems. • First design of MMC-based AGPS for the DEMO NBI with numerical simulations. • Space evaluation of MMC-based solution and comparison with ITER AGPS. The present conceptual design of the Neutral Beam Injectors (NBIs) for DEMO foresees an acceleration voltage of 1 MV, subdivided in 5 steps of 200 kV each, as for the ITER NBI. This voltage is provided by the Acceleration Grid Power Supply (AGPS), which shall also guarantee the capability of handling repetitive grid breakdowns and provide the required high dynamic response, low ripple and high accuracy. These requirements addressed the ITER AGPS design towards a challenging solution based on multiple conversion stages, where the last rectification and filtering stage is gas-insulated. The research here presented aims at evaluating the possible utilization of Modular Multilevel Converters (MMCs) for the AGPS of future NBIs. A preliminary design of the DEMO AGPS has been carried out, where each 200-kV voltage step is generated by an air-insulated MMC stage operating as a rectifier. The main circuit parameters of the MMC have been determined, considering normal and anomalous working conditions. A numerical model has been developed to verify the converter performances in stationary and dynamic conditions and to simulate breakdown and beam-off events. The area occupied by the proposed MMC-based AGPS has been tentatively estimated and compared to that of the ITER solution. The results of the analysis are shown in this paper. Finally, strengths and criticalities of the MMC solution with respect to the ITER-like approach are discussed. [ABSTRACT FROM AUTHOR]

• Features of 2D and 3D MHD flows typical in liquid metal blankets are reviewed together with available correlations for MHD pressure losses. • Design MHD guidelines that support a preliminary estimate of MHD effects in a blanket concept, in terms of pressure drop and flow distribution, are discussed. • Considerations on MHD issues for blanket applications focus on present status of the analysis of major problems and on future research needs. • Description of experimental methodology to be followed in order to produce experimental data for code validation purposes. The design of breeding blankets represents the major challenge for fusion reactor engineering because of performance requirements and severe operating conditions in terms of heat load and neutron flux. Liquid metal alloys such as lead-lithium, due to their lithium content, can be used to breed tritium, one of the plasma fuel components, and owing to their high thermal conductivity, they may serve as coolants. On the other hand, there are technical issues related to the fact that the liquid metals are electrically conducting and interact with the plasma-confining magnetic field. Induced electric currents and generated electromagnetic forces affect velocity and pressure distribution in the blankets. Magnetohydrodynamic (MHD) flows for fusion applications have been often investigated in simplified geometries, such as pipes, ducts, bends, with focus on their fundamental features. These analyses are essential, since results remain valid as background for the development of blanket designs, even when a concept is dismissed. However, the conceptual study of fusion blankets requires to take into account the global multiple effects, that arise when the full system is considered. Progress made in fusion-related MHD research results from combined numerical and experimental activities. In this paper we review and summarize features of 2D and 3D MHD flows that are typical in liquid metal blankets, together with available correlations for MHD pressure losses. This knowledge can provide simple design MHD guidelines that support a preliminary estimate of MHD effects in a blanket concept, in terms of pressure drop and flow distribution. [ABSTRACT FROM AUTHOR]

• The 3Sm rule, Bree diagram and Efficiency index diagram have been studied and compared, indicating that 3Sm rule is over conservative, the conservative between Bree diagram and Efficiency index diagram depends on the ratio of primary stress to yielded stress. • The elastic-plastic structural analyses are performed to compare and verify the results from the elastic structural analysis. The plastic analysis results indicate that the 3Sm is too conservative for the FW design, and the Efficiency index diagram is not conservative. • The Bree diagram considering the impact of the secondary membrane stress (ASME BPVC VIII) is consistent with the elastic plastic results, which can be an option for the ratcheting assessment. This paper focuses on the structural design criteria (3Sm, Bree Diagram and Efficiency Index Diagram) to prevent ratcheting damage for fusion component design, which includes conservative comparison of shakedown / ratcheting evaluation criteria and the elastic / elastoplastic analysis of first wall. The results show that: 3Sm rule is over conservative for low primary stress component while is less conservative than Bree Diagram for high primary stress (~ yield stress Sy) component. The conservatism between Efficiency Index Diagram and Bree Diagram is different with the specific structural primary stress. According to the elastic and elastoplastic analysis of the first wall, the 3Sm rule is too conservative, which may decrease the economy of the first wall. However, common Bree Diagram and Efficiency Index Diagram in nuclear pressure vessel design criteria (ASME BPVC III, SDC-IC and RCC-MRx) cannot effectively ensure the safety of the first wall. The mainly reason is that the mechanical characteristics of high thermal membrane stress in the first wall is different with fission pressure vessels. The modified Bree Diagram in ASME BPVC VIII, considering the impacts of thermal membrane stress on ratcheting, meet with the nonlinear result and is more reasonable. As a result, considering economy and conservatism of the first wall, it is suggested that Bree Diagram of ASME BPVC VIII is the most suitable shakedown / ratcheting assessment rule in the design of first wall though ASME BPVC III is more commonly used for nuclear pressure vessels. [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]

• A double null architecture could allow a wider range of remote maintenance options. • With a split blanket design there are more service pipes exiting the lower port. • Inserting the cutting/welding tool outside the outboard port wall is advantageous. • Pipe design must allow for remote manipulation of divertor and blanket segments. • Using large radii pipes utilises port space and has remote maintenance advantages. A study into the viability of a double null concept for DEMO has been undertaken as part of the work to address Key Design Integration Issues within the DEMO powerplant. The objective of this study was to evaluate the feasibility of a double null architecture featuring split blanket segments to enable a wider range of maintenance options. One of the main challenges found was the integration of the service pipes to provide connections to the cooling and tritium breeding systems for the divertor and breeding blanket. By understanding how these pipes might be integrated, the constraints on key details such as the length of the lower port could be understood. This allowed for progression of other areas of the design and investigations into the remote handling of these pipes during maintenance periods. The key requirements identified for the service pipe routing included determining the removal sequence for the blanket and divertor segments during maintenance; access for the cutting and welding tool at the pipe/blanket interface; pipes grouped as modules per component; and optimisation of port space for the vacuum pumps. Two proposals were considered to investigate their impact on the insertion location for the cutting and welding tool: pipes with large radii, and pipes with smaller radii. The former was found to be preferable due to the remote maintenance advantages, versus the limited space gains offered by the latter. This paper will describe the development of these requirements, both proposals, and discuss the challenges uncovered during the study in more detail. [ABSTRACT FROM AUTHOR]

Two models are currently taken as reference for the breeding blanket (BB) of DEMO, the Helium Cooled Pebble Bed (HCPB) and the Water Cooled Lithium Lead (WCLL). For both the models two Balance of Plant (BOP) configurations are currently investigated. They are all based on four independent primary heat transfer systems (PHTSs). The largest PHTS is devoted to remove the BB thermal power, two PHTSs for the divertor (DIV) and one for the vacuum vessel (VV) heat removals. Both DIV loops and VV loop transfer their thermal power directly to the power conversion system (PCS). The two BOP configurations are investigated to cope with the pulsed operation of the reactor. The first configuration consists of an "indirect coupling" of the BB PHTS to the PCS by an Intermediate Heat Transfer System placed in between them. The second configuration consists in a "direct-coupling" of the BB PHTS to the PCS. Reliability, availability, maintainability and inspectability (RAMI) analyses for the four PHTS and PCS configurations are presented in this paper. The Failure Mode, Effect Analysis (FMEA) and the Reliability Block Diagram (RBD) methodologies have been used to perform the studies. As main result can be highlighted that even the low reliability of the systems to operate without failure and the large number of requests of corrective maintenance, all the systems are in compliance with an operational availability target of 30 %. [ABSTRACT FROM AUTHOR]

Tritium permeation from breeder material to the Water Coolant System (WCS) in Water Cooled Lithium Lead (WCLL) Breeding Blanket (BB) is one of the technological issues to be solved in the design of the European DEMO. Since the tritium extraction from the Water Coolant System is more challenging and expensive than the extraction from the eutectic alloy PbLi, it is mandatory to use of a protective coating on the blanket wall to minimize the permeation rate. Moreover, a protective coating can prevent the corrosion of EUROFER steel by the action of PbLi. alumina-based coatings are considered as reference for barriers thanks to their good chemical compatibility with the PbLi alloy and their capability to reduce permeation. Three coating technologies were selected in the frame of the EUROfusion project: electrochemical ECX (chemical deposition) process, Pulsed Laser Deposition (PLD) and Atomic Layer Deposition (ALD) coating. The coatings were developed and optimized in order to satisfy the design requirements of good mechanical compatibility with steels, strong adhesion, corrosion compatibility in PbLi at relevant BB design conditions and a Permeation Reduction Factor at least of 200 under neutron irradiation. The present paper aims to describe the status of the technologies and the main results obtained. The final objectives of the R&D activities are to demonstrate the applicability of the coating to WCLL BB and therefore the scale-up of the technologies from laboratory scale to the BB scale. [ABSTRACT FROM AUTHOR]

• First neutronics assessment of alternative divertor configurations in DEMO. • Shielding proposals to reduce TF coil heat loads and neutron streaming. • DPA, helium production and to assess material damage and lifetimes of components. • MCNP model constructed for SN, SX and XD divertor configurations. • Tritium breeding ratio (TBR) calculated for SN, SX and XD divertor configurations. As a demonstration fusion power plant, EU DEMO has to prove the maturity of fusion technology and its viability for electricity production. The central requirements for DEMO rest on its capability to generate significant net electric power to the grid (300 MW to 500 MW) safely and consistently. Plant availability and lifetime will approach that of a commercial fusion power plant. Operating at such regimes presents many complex challenges, of which one is plasma exhaust. To mitigate the risk that the implementation in preceding experimental devices, namely ITER, does not extrapolate to the requirement of DEMO, alternative solutions must be sought. The investigation of alternative divertor configurations was born out of this motive, seeking to resolve a 'critical' challenge for the realisation of DEMO. In this paper, we study the neutronics performance of three concepts: Single Null (SN), Super-X (SX) and X-divertor (XD). This is the first time a preliminary analysis of alternative configurations to the SN baseline has been performed. The shielding proposals and design recommendations presented herein should be integrated with other engineering and physics constraints in future iterations of the chosen divertor concept. [ABSTRACT FROM AUTHOR]

The European Research Roadmap to the Realisation of Fusion Energy foresees that the DEMO reactor is going to succeed ITER in the pathway towards the exploitation of nuclear fusion, achieving long plasma operation time, demonstrating tritium self-sufficiency and producing net electric output on an industrial scale. Therefore, its design must be more oriented towards the Balance of Plant (BoP) than it is in ITER. Since the early pre-conceptual phase of the DEMO project, emphasis has been laid on identifying the main requirements affecting the overall architecture of the BoP. For instance, specific efforts and proper solutions have been envisaged to cope with the pulsed nature of the heat source. Furthermore, the current development of two blanket concepts calls for two separate BoP options to be conceived. This paper summarizes the main alternatives outlined at the end of DEMO pre-conceptual design phase for the BoP concepts based on both the Helium-Cooled Pebble Bed (HCPB) and Water-Cooled Lithium Lead (WCLL) Breeding Blanket (BB) technologies. Then, the assumed reference configurations of both the BoP concepts are described in detail, highlighting the main features and the most relevant engineering aspects. Attention will be focussed on technological challenges, integration constraints and other open issues, highlighting pros and cons of the chosen BoP options to be further investigated in the next design phase. [ABSTRACT FROM AUTHOR]

In the framework of the EU DEMO fusion reactor technology development, a new steam generator consisting in a helical tube bundle is currently under study for the lithium-lead loop of the Dual Coolant Lithium Lead (DCLL) and Water Cooled Lithium Lead (WCLL) breeding blankets. This solution turns out to be very interesting for high power plants, since the helical geometry is very compact and it assures a high thermal power exchanged, taking up the minimum amount of space. In this framework, the ENEA Brasimone Research Centre supports the development of this innovative component by means of experimental activities, involving CIRCE, a large scale pool-type facility using Lead-Bismuth Eutectic as primary fluid and pressurized water as secondary fluid. The main components of a new test section named THETIS are: a Fuel Pin Simulator, acting as primary heating source, a vertical mechanical pump for the primary fluid circulation inside the main pool and a new prototypical helical coil steam generator mock-up, designed to be relevant for the DCLL and WCLL lithium-lead loop, acting as primary heat sink. In such configuration, the facility will be involved in a set of tests aiming at demonstrating the technological feasibility and thermal-hydraulic performances of this prototypical steam generator, as well as the suitability of the component for the lithium-lead loop in DEMO. The experiments will also provide a database for system thermal-hydraulic codes validation. The aim of this paper is to present the main layout of the CIRCE facility, to describe the preliminary design of the test section and the main features of the helical coil steam generator mock-up. Furthermore, a preliminary test analysis carried out by the system thermal hydraulic code RELAP5/Mod3.3 is presented. A numerical 1-D model of the helical coil steam generator has been set-up in order to test the performance of the component from a thermal-hydraulic point of view. An additional 3D CFD analysis was performed for the flow field of the HCSG to assess the secondary flow behaviour in the component and the pressure losses. [ABSTRACT FROM AUTHOR]

In the frame of the activities promoted and encouraged by the EUROfusion Power Plant Physics and Technology (PPPT) department aimed at developing the EU-DEMO fusion reactor, strong emphasis has been recently addressed to the whole Balance of Plant (BoP) which represents the set of systems devoted to convert the plasma generated thermal power into electricity and to deliver it to the grid. Among these systems, the Divertor Primary Heat Transfer Systems (PHTSs) are intended to feed coolant to the two main components of the Divertor assembly, namely the Plasma Facing Components (PFCs) and the Cassette Body (CB). Since the DEMO Divertor must withstand high heat flux loads together with a considerable neutron deposited power, very tight tolerances may be allowed to the coolant inlet conditions. Therefore, the design of reliable PHTSs is of the utmost importance towards the development of an EU-DEMO fusion reactor. Within this framework, a study has been jointly carried out by University of Palermo, Ansaldo Nucleare and CREATE to design the Ex-Vessel PHTSs of both the PFCs and the CB for a DEMO reactor equipped with a Helium Cooled Pebble Bed Breeding Blanket. The paper describes criteria and rationale followed with the aim to achieve simple PHTS designs based on the adoption of easy-to-manufacture main components avoiding too much extrapolation from the state-of-art technology. Results of preliminary thermal-hydraulic calculations carried out to size heat exchangers, pressurizers, piping and pumps are presented and critically discussed, with particular attention to those integration, safety and feasibility constraints that may deeply affect the design of such components. Finally, the evaluation of PHTS key parameters as total pumping power and coolant inventory is reported. [ABSTRACT FROM AUTHOR]

In the framework of the DEMO Work Package Balance of Plant of EUROfusion consortium, ENEA has been involved in experimental and numerical activities related to the development of a prototypical heat exchanger, suitable as steam generator for the lithium-lead loop of the Dual Coolant Lithium Lead and Water Cooled Lithium Lead breeding blankets. For this purpose, an experimental campaign has been executed on the pool-type liquid metal-cooled facility CIRCE at ENEA Brasimone Research Centre. A dedicated test section named HERO has been designed and installed inside the main vessel of CIRCE. The innovative steam generator consists of a tube bundle with seven double walled bayonet tubes, fed with pressurized water. The selected configuration improves the plant safety, reducing the possibility of water-lead/lead-alloy interaction thanks to a double physical separation and allowing an easier control of eventual leakages from the coolant by pressurizing the separation region with inert gas. A set of tests has been defined to demonstrate technological feasibility and performances of this prototypical steam generator, as well as the suitability of the component for the lithium-lead loop in DEMO. In particular, one of the performed tests is presented and discussed in this paper. The experiment is characterized by a secondary loop feedwater working pressure of 10 MPa and a steam generator inlet temperature of 300 °C. On the primary side, the lead-bismuth eutectic has been used as working fluid with a steam generator inlet temperature of 480 °C. During the test, an experimental sensitivity analysis on the primary coolant mass flow rate has been performed. Furthermore, the results of a post-test analysis realized with two versions of the system thermal-hydraulic code RELAP5 are presented, in order to evaluate their capability in simulating the performances of the component and to support the validation process of the codes for heavy liquid metal applications. The work is concluded presenting a scaling analysis to find the equivalence between LBE and PbLi, recalculating the available experimental data with RELAP5 code using PbLi as working fluid. [ABSTRACT FROM AUTHOR]

Spectroscopic measurements have been recently proposed in DEMO for divertor detachment control. In its current design, the DEMO Divertor Survey Visible High-Resolution Spectrometer is foreseen to perform spectroscopy measurements by integrating three optical subsystems into an equatorial port (EP). Behind the first wall, light travels through a set of metallic mirrors and ducts before it reaches the closure plate of the EP. This paper presents a nuclear analysis performed with the Monte Carlo simulation program MCNP6 for two alternative configurations of the system. The results show that the configuration with 5 mirrors per transmission line is very effective to reduce the neutron streaming through the port. However, it will not be possible, with the current design, to introduce standard electronics along the spectroscopy ducts, as the dose rate limits for non-critical electronic components are exceeded in both configurations. In the plasma-facing mirrors, the heat loads are below 2 mW/cm3, which shows that the strategy of recessing the first mirrors and placing them behind small-diameter openings is effective to decrease the loads in the mirrors. FISPACT simulations for different materials show that material transmutation in the mirrors will be negligible throughout the DEMO reactor lifetime. [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]

DEMO is a key part of the EU fusion roadmap, where the programme is reaching the end of the pre-conceptual phase with a gate review in 2020. As part of the work to complete this phase, eight Key Design Integration Issues (KDIIs) have been identified as critical to the programme [ 1 ] (Bachmann, 2020). Within KDII#1 (Wall protection to withstand plasma transients) the feasibility of the Inner Mid-plane Limiter (IML) is assessed from a remote maintenance perspective. The IML is an actively cooled component attached to the inner ring of the tokamak torus, with access for removal and installation only possible from the plasma facing side. The IML service life is less than that of the Breeder Blankets (BBs), due to the foreseen transients and/or the CuCrZr cooling pipes that have limited irradiation lifetime. This drives the need to change the IML with the BB in position. When replacing the IML, the pipework is expected to be too highly irradiated to allow re-welding. This drives the need to change the pipework to the IML, a challenging task when the BB are in their installed position. This paper presents the preliminary development of a maintainable IML concept, including: the development of the IML fastening for remote maintenance; a proposal for a new IML cooling pipe chute; and the rationale for the options selected. [ABSTRACT FROM AUTHOR]