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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]

Within the DEMO first wall 3D shape design activity, studying the effect of misalignment started in 2017. Such assessments have been conducted in the past for ITER and heat flux penalty factor maps have been created [ 1 ]; this route could be a feasible approach in the case of DEMO as well. This paper details the methodology that allows the effects of misalignments to be assessed for DEMO. The test cases focus on steady-state plasma operation (start of flat top). The aim is to understand the effect of basic misaligned cases, for example radial protrusion/recession or poloidal rotation of a single module. To do so, particle tracing software codes such as SMARDDA and PFCflux have been used to create heat flux maps that reach the first wall surfaces. These heat flux maps, combined with the specified radiative heat load, are used as input for simplified FE models of the blanket modules. As a result, not only the effect on heat flux, but also on the temperature (and later stress) distributions can be estimated. The paper describes how the obtained results can be implemented in ANSYS in the identified critical cases from the test matrix that has been studied. The results obtained from the nominal heat flux map are compared to the misaligned cases. The mitigating effect of the 3D nature of the heat conduction on the peak temperature is discussed. This work paves the way to assessing more realistic combined misaligned cases (such as misalignment from different thermal expansion, or due to electromagnetic loads etc. of neighbouring blankets) in the future. [ABSTRACT FROM AUTHOR]

This paper presents the analyses performed within the SAE (Safety and Environment) project of EUROfusion/PPPT aimed at providing up-to-date estimates of the activity inventories and the decay heat generation in the WCLL. Detailed evaluation of the WCLL PPPT reactor concept based on the generic Multi-Module-Segment (MMS) model was performed by calculating the decay heat and activity inventories for the WCLL central inboard/outboard blanket modules (DEMO 2015 baseline, MMS design). Following, the estimation of radial distribution of decay heat and activity in breeder blanket module material mixture is presented. And finally, the preliminary approach to evaluate Li flow inside the breeder zone is presented in this paper. Additional calculations were performed to identify dominant nuclides after the irradiation. [ABSTRACT FROM AUTHOR]

The roadmap to the realization of fusion energy describes a path towards the development of a DEMO tokamak reactor, which is expected to provide electricity into the grid by the mid of the century (Romanelli, 2013). The DEMO diagnostic and control (D&C) system must provide measurements with high reliability and accuracy, not only constrained by space restrictions in the blanket, but also by adverse effects induced by neutron, gamma radiation and particle fluxes. In view of the concept development for DEMO control, an initial selection of suitable diagnostics has been obtained (Biel et al., 2019). This initial group of diagnostic consists of 6 methods: Microwave diagnostics, thermo-current measurements, magnetic diagnostics, neutron/gamma diagnostics, IR interferometry/polarimetry, and a variety of spectroscopic and radiation measurement systems. A key aspect for the implementation, performance and lifetime assessment of these systems on DEMO, is mainly attributable to their location, that must be well protected, and meet their own set of specific requirements. With this in mind, sightline analysis, space consumption and the evaluation of optical systems are the main assessment tools to obtain a high level of integration, reliability and robustness of all this instrumentation; essential features in future commercial fusion power nuclear plants. In this paper we concentrate on spectroscopic and radiation measurement systems that require sightlines over a large range of plasma regions and inner reactor surfaces. Moreover, this paper outlines the main results and strategies adopted in this early stage of DEMO conceptual design to assess the feasibility of this initial set of diagnostic methods based on sightlines and the integration of these needed for DEMO D&C. [ABSTRACT FROM AUTHOR]

In this paper we present the methodology implemented for analyzing System Requirements and Interfaces of the Heating and Current Drive (HCD) system of the European Demonstration Fusion Power Reactor DEMO. The work consisted in updating the preliminary framework of the Model-Based Systems Engineering model of the HCD System Architecture. This is now containing an ontology, a set of 6 perspectives and a defined set of viewpoints for each Perspective, for refining the HCD System Architecture. The scope of the work is to manage the interdependencies of HCD system elements and their integration into DEMO, for a given set of system functions. On the one hand, this means to address the identification and definition of the interfaces occurring, both internally in the HCD system, and between the HCD system and neighboring systems. On the other hand, this implies studying the impact of requirements coming from the ongoing physics studies. The rationale is to provide the technical foreground for supporting the decision-making processes related to the HCD system which is planned to be carried out during the forthcoming Conceptual Design Phase. The results we show in this paper are part of the design and integration activities consisting of both systems engineering methodologies and design analysis, all aiming at ensuring consistency in the overall EU DEMO plant design. In this framework the DEMO Heating and Current Drive system has been selected as pilot project for the application of Systems Engineering methodologies. [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]

Highlights • The performance and reliability of the breeding blanket are foremost considerations in the development and deployment of fusion power. • No breeding blanket has ever been built or tested. Hence, its integrated functions and reliability represent a very challenging development. • This paper provides an overview of the revised design and development strategy for the DEMO breeding blanket in Europe. • It includes a proposal to change the EU TBM options to be tested in ITER, in particular to replace a HCLL concept with a WCLL concept. • This is needed to obtain important information from the two current breeders (solid and liquid) and coolants (helium and water) being considered for DEMO. Abstract This paper provides an overview of the newly revised design and development strategy for the DEMO breeding blanket in Europe. This has been defined to take into account the input from the DEMO pre-conceptual design activities, the findings and recommendations of a thorough technical and programmatic assessment of the breeding blanket and the EU ITER Test Blanket Module (TBM) programs, conducted in 2017 by an independent expert panel. This work has led to the identification of (i) the most mature and technically sound breeding blanket concepts to be used as "driver" and "advanced" breeding blankets in DEMO, the latter to be installed and tested in a limited number of properly designed segments, potentially being more attractive for future fusion power plants; (ii) the remaining technical gaps and R&D priorities. A number of urgent steps that are required to better align and strengthen the EU TBM and DEMO Breeding Blanket Program as a whole and to aim at an efficient implementation of the work are described in this paper. These include a proposal to change the EU TBM options to be tested in ITER in order to obtain important and useful information from the two current breeders (solid and liquid) and coolants (helium and water) considered for DEMO. [ABSTRACT FROM AUTHOR]

Highlights • A CO 2 -cooled FW has been proposed in this paper. • A high heat flux with radiation heat flux and particle heat flux has been applied on FW. • Different wall roughness position of channel has been shown. • Different wall roughness value has been given. • Sensitivity particle heat flux has been studied in this paper. Abstract The Demonstration Fusion Reactor in Europe (EU-DEMO) is considered to be the last step before building a commercial fusion power plant. Up to now, He has been chosen as gas coolant for three of the four candidate breeding blankets proposed for the EU-DEMO, being the Helium Cooled Pebble Bed (HCPB) among them. The choice of He coolant is supported, among other features, by its superior heat transfer capabilities, its transparency to neutrons and chemical inertness. However, the use of this coolant in DEMO poses some feasibility questions, like the large coolant inventory, resulting in possible safety issues, as well as the large circulating power. On the other side, CO 2 has been the fluid coolant choice in the nuclear fission industry for gas-cooled reactors since the 1950s. The key advantage of this gas is its larger density, which results in more compact components, lower coolant inventories, lower circulating powers and thus better plant thermal efficiency. This, coupled also with a relatively high molecular stability at moderate temperatures (up to 600 °C), its larger availability and lower price make CO 2 an attractive coolant for a pebble bed breeding blanket, as an alternative to He. As a Plasma Facing Component, the First wall (FW) is one of the components under the most challenging environment. Currently, the FW region at the upper port inboard location is expected to suffer the most critical heat flux loads. Indeed, a peak charged particle heat flux 0.99 MW/m2 together with a radiation heat flux of 0.18 MW/m2 is foreseen at this location. This paper reports the thermal-hydraulic analyses carried out to find the practical heat flux limits by using CO 2 as coolant. The analyses have been performed by using the commercial computational fluid dynamics (CFD) code ANSYS CFX. Different augmented technical wall roughness areas have been considered, for a best heat transfer-to-pressure drop ratio. Sensitivity analyses have been performed as well for higher peak particle heat fluxes, evidencing that a CO 2 -cooled FW can dissipate peak heat fluxes up to 1.8 MW/m2 keeping the circulator power in a reasonable range. [ABSTRACT FROM AUTHOR]

Highlights • An improved geometrical conception of the DEMO Divertor Cassette has been considered and a suitable finite element mesh has been adopted. • The loads really acting on this component have been evaluated detailly and applied overcoming the early hypothesis of uniform distribution: so the load spatial gradients have been reprioduced with more precision. • The obtained results confirm the previous deductions: the main role has been carried out by the thermal loads, and moreover the electric-magnetic loads determine a state of stress far lower than that caused by the thermal loads. • These thermal loads are the results of a dedicated thermal-fluid–dynamic numerical code properly developed for this aim. • Also in this case, an early application of RCC-MRx design rule have been applied: the Divertor Cassette results to be safe in nominal conditions while the same assessment states that the knuckle stress state is above faliure. Abstract This paper presents the enhancements related to the structural analyses of DEMO Divertor in the framework of the EUROfusion Power Plant Physics & Technology (PPPT) program. This activity started two years ago and its preliminary results were published in previous papers. It has been divided in some areas defined by the similarity of the matters they contain: the structural analysis, of utmost importance, has been preceded by a preliminary phase, like the geometry definition or the thermal and the electric-magnetic analysis for loads evaluation; then the structural analysis has been finally confirmed with further evaluations related to excessive deformation or plastic instability. This paper discusses the improvements adopted either in the preliminary analysis or in the main structural analysis. Specifically it highlights the introduction of the thermal and electro-magnetic loads application in terms of a detailed spatial distribution that is now available. More the supports have been introduced in the model and their structural behavior has been investigated considering their interaction with the cassette. The structural assessment, according to the Design and Construction Rules for Mechanical Components of Nuclear Installation (RCC-MRx), has been performed either for the cassette or for the attachments: for the former it is positive while for the latter case serious limitations have been found. [ABSTRACT FROM AUTHOR]

Highlights • The balance of plant of DEMO have been described in this paper. • A new kind of coolant for solid breeder blanket of CCPB have been proposed. • Comparative analysis of the efficiency of HCPB and CCPB for DEMO have been given. • Parametric analyses of BB outlet temperature and CCPB pressure drop have been shown. Abstract The EU DEMO is considered to be the last step before building a commercial fusion power plant. Up to now the Helium Cooled Pebble Bed (HCPB) Breeding Blanket (BB) is one of the four BB concepts being investigated in the EU for their possible implementation in the EU DEMO. While He is the reference gas cooling for the BBs in DEMO, CO 2 has been the coolant choice in the nuclear fission industry for gas-cooled reactors during the past 60 years. In the recent years, it is also gaining interest as a secondary coolant for the Balance of Plant (BoP) to replace the traditional vapor (Rankine) cycle. Despite the lower heat capacity and thermal conductivity of CO 2 with respect to He, the larger density of the former can lead to an advantageous performance on heat dissipation rate per unit of pumping power. This feature, coupled with its relative transparency to neutrons, the molecular stability at the operating temperatures of the BB and the industrial scale ready turbomachinery makes CO 2 an attractive alternative to He. The aim of this paper is to make a parametric study of the gross and net efficiencies and the design of a simplified BoP of a DEMO power plant based on a CO 2 -cooled compared to the reference EU DEMO HCPB BB, as a function of the BB outlet temperature and turbine inlet pressure. From this study it is concluded that the vacuum vessel and the divertor heat sources are benefitial in order to improve the net plant efficiency for both coolants. Also, it has been observed that the plant net efficiency is better for the case of a PHTS with CO 2 as heat transfer fluid under the assumption that both BBs share the same geometry. In the case that the BBs do not share the same geometry, the study concludes that a PHTS with CO 2 leaves a pressure drop margin of about 1.5 before reaching the same net efficiency as a HCPB BoP DEMO plant. [ABSTRACT FROM AUTHOR]

Abstract In the framework of the European "HORIZON 2020" innovation and research programme, the EUROfusion Consortium develops a design of a fusion power demonstrator (DEMO). One of the key components in the fusion reactor is the Breeding Blanket (BB) surrounding the plasma, ensuring tritium self-sufficiency, heat removal for conversion into electricity, and neutron shielding. CEA-Saclay, with the support of Wigner-RCP and Centrum výzkumu Řež, is in charge of the development of one of the four BB concepts investigated in Europe for DEMO: the Helium Cooled Lithium Lead (HCLL) BB. The rationales of the HCLL are the use of Eurofer as structural material, eutectic liquid lithium-lead (PbLi) as tritium breeder and neutron multiplier, and helium gas as coolant. This paper shows the basic description of the DEMO HCLL BB concept and its design evolution during the past years, from a design based on the ITER Test Blanket Module (TBM) concept to a more advanced design called "Advanced-Plus" concept. This new HCLL BB concept that has been designed in order to improve Tritium Breeding Ratio (TBR) and shielding performances is presented. This new reference HCLL BB design has been analyzed and show very promising nuclear performances. Nevertheless, the "Optimized Conservative" concept, based on ITER TBM, is still considered as a robust back-up solution since structural improvements are still necessary on the "Advanced-Plus" concept. Moreover, a new Back Supporting Structure (BSS) is presented in this paper, designed to support the BB modules, with the aim to reduce pressure drops and thermal stresses. [ABSTRACT FROM AUTHOR]

Highlights • Conceptual design of irradiation PbLi capsule has been finished. • Numerical analyses were performed to support capsule design. • Two concepts of temperature control system are proposed. • Out-of-pile experiments are planned for performance testing and design validation. Abstract Three concepts of the European DEMO Breeding Blankets currently under development within the EUROfusion project use as a breeder eutectic lithium-lead (PbLi). Since the inner walls of the blanket modules and the associated ancillary systems has to withstand severe corrosion attack, protective aluminum-based corrosion and tritium permeation barriers are being developed within the Breeding Blanket project. The goal of this activity is to evaluate properties of these protective coatings in conditions of a combined effect of hot PbLi and neutron irradiation from the LVR-15 reactor. The reduction effect of coatings on tritium permeation will be measured comparing the tritium permeation rate between coated and uncoated Eurofer samples. The paper presents the major results of the preliminary neutronics and thermal-hydraulics calculations leading to the proper design of the irradiation capsule which is also described in this paper. [ABSTRACT FROM AUTHOR]

Abstract The role of nuclear data for neutronics analyses of Fusion Technology (FT) facilities is presented in this paper. Nuclear data evaluations of high quality are required for transport simulations, uncertainty assessments, activation and radiation damage calculations affecting the design and performance of the FT facilities, as well as safety, licensing, waste management and decommissioning issues. A corresponding programme on the nuclear data development (NDD) and qualification is conducted within the Power Plant Physics and Technology (PPPT) programme of EUROfusion supporting the development of the DEMO fusion power plant and the IFMIF-DONES neutron source. The programme builds on the achievements of pre-ceding NDD activities within the European fusion programme as addressed in the paper. Further needs for design, shielding, activation and radiation dose calculations are discussed, and recommendations are given to further improve and qualify the nuclear data base for PPPT nuclear analyses. This includes dedicated experimental activities tailored to the needs of DEMO and IFMIF-DONES. [ABSTRACT FROM AUTHOR]

The present paper addresses the impact of the divertor option on the nuclear performances of the Demonstration fusion reactor (DEMO). As the effect of the number and size of the divertor has been already evaluated, in this work the focus has been posed on the composition in terms of amount of cooling inside the divertor cassette. Transport responses, as the Tritium breeding ratio (TBR), neutron and gamma fluxes and spectra inside the plasma chamber, as well as activation responses such as shutdown dose rate, decay gamma fluxes and heating have been evaluated for two different blanket concepts of the future European DEMO reactor: DCLL and WCLL. Three different divertor compositions have been tested demonstrating the importance of this component not only locally but in the global radiation field. The transport analysis has been performed with the Monte Carlo code MCNP5 and the JEFF3.1.1 and JEFF3.2 data libraries. The activation responses calculated using Advanced D1S method have been recently assessed and summarized in the present paper. [ABSTRACT FROM AUTHOR]

The EUROfusion Consortium develops a design of a fusion power demonstrator plant (DEMO) in the framework of the European “Horizon 2020” innovation and research program. One of the key components in the fusion reactor is the Breeding Blanket (BB) surrounding the plasma, ensuring tritium self-sufficiency, heat removal for conversion into electricity, and neutron shielding. Among the 4 candidates for the DEMO BB, 2 of them use helium as coolant (HCPB, HCLL), and another one (DCLL) uses helium to cool down the First Wall (FW) only. Due to uncertainties regarding the plasma Heat Flux (HF) load the DEMO BB integrated FW will have to cope with, a set of sensitive thermal and stress analyses have been performed in order to define the possible margin against HF the integrated Helium-Cooled Eurofer FW could have. Based on the Helium Cooled Lithium Lead (HCLL) equatorial outboard module dimensions, thermal and stress Finite Element Method analyses have been performed with Cast3M with various FW front wall thicknesses and HF, under normal steady state condition. Stress have been analyzed with RCC-MRx code including high temperature (creep), cyclic (fatigue) and irradiated rules. This paper shows that the thickness of the plasma-facing wall of the FW should be minimized, within the limits necessary to withstand primary stresses, in order to reduce the temperature on the structure and thus prevent fatigue and creep damage as well as a reduction of the stress limits Sm, function of temperature, to prevent ratcheting. Moreover, the paper will discuss the importance of having constant HF during the reactor operation. A small variation of HF could increase a lot the risk of damage such as fatigue and creep. At the end, the effect of irradiation shows up to be the limiting criterion and penalizes the capacity of the FW to withstand high HF. [ABSTRACT FROM AUTHOR]

For ITER or the future DEMO remote maintenance system (WPRM), several types of special tailored automatic manipulators are needed for vacuum vessel (VV) component transportation, inspection, and removal from and replacement to the VV wall. Due to the extreme work conditions such as big size, big payload and high environment temperature, both the static and dynamic error sources should be considered when design a controller. In this paper, the accuracy improvement issues regarding the static and dynamic error sources are taken into account. The paper proposed a plane constraint calibration method for static error calibration, whilst the dynamic errors are calibrated using a linear inverse dynamic error model. Differential Evolution global optimization method was used to generate optimal excitation trajectories, and to identify the static and dynamic parameter errors. A 6-DOF (degrees of freedom) commercial industrial serial robot was employed to carry out the corresponding simulation and experimental studies. This paper mainly focuses on the general case studies, the results found in this research would be extrapolated to support the ITER or the future DEMO remote handling systems. [ABSTRACT FROM AUTHOR]

For the DEMO torus vacuum system, liquid ring pumps (LRPs) with mercury as working fluid have been proposed. These vacuum pumps shall be used for pumping the torus down to a pressure <100 Pa and for providing rough vacuum for the torus primary pumps. Unfortunately, liquid ring pumps operated with mercury have never been built and tested before, so no operational experience is available. It is not even clear if the pump will work with such a high density working fluid or if the pump can withstand the high mechanical forces. Therefore, proof-of-principle-testing has become necessary. This is the very first step in the development of a DEMO relevant LRP that is hermetically tight and fully tritium compatible. The paper describes the proof-of-principle experiments that have been performed in the THESEUS pump test facility at KIT and its results. Special focus of the paper is given to operational aspects as well as to mercury handling procedures. The impact of these findings on the design of future LRPs in a DEMO relevant design is shown up and modifications of the existing design are proposed. [ABSTRACT FROM AUTHOR]

This paper refers to the activity of structural design of DEMO Divertor in the framework of the EUROfusion Consortium. The structural analysis and its preparatory assessments were carried on since a year and the first results were published in a previous paper. The Cassette Body has been examined considering the most conservatives loads (e.g. coolant pressure, volumetric nuclear heating and electro-magnetic loads) according to their latest estimates. This work is based on the design-by-analysis approach adopted in the conceptual design phase of the DEMO Divertor. This design activity has been focused on some key parameters e.g. loads, main geometric dimensions, positions of the Cassette attachments on the vacuum vessel, way of loads application to characterize the structural behavior of the Divertor Cassette. In addition to the existing 3D solid element model, a shell element model has also been developed: with this new model a parametric analysis can be done for a fast optimization. The structural assessment was done according to the Design and Construction Rules for Mechanical Components of Nuclear Installation (RCC-MRx). [ABSTRACT FROM AUTHOR]

Since 2014 preconceptual design activities for European DEMO divertor have been conducted as an integrated, interdisciplinary R&D effort in the framework of EUROfusion Consortium. Consisting of two subproject areas, ‘Cassette’ and ‘Target’, this divertor project has the objective to deliver a holistic preconceptual design concept together with the key technological solutions to materialize the design. In this paper, a brief overview on the recent results from the subproject ‘Cassette’ is presented. In this subproject, the overall cassette system is engineered based on the load analysis and specification. The preliminary studies covered multi-physical analyses of neutronic, thermal, hydraulic, electromagnetic and structural loads. In this paper, focus is put on the neutronics, thermohydraulics and electromagnetic analysis. [ABSTRACT FROM AUTHOR]

In the present pre-conceptual design, the base converters to supply the main DEMO poloidal magnets are rated for 45 kA and about 10 kV. If the traditional design approach, based on thyristor bridges, was adopted, this would result in very large reactive power exchanged when low voltage values are required by the load. To satisfy the limitations imposed by National Grid Operators, the reactive power should be compensated inside the DEMO Plant Electrical System by means of a Reactive Power Compensation and Harmonic Filtering system. This could represent an additional cost and takes up a large area of the plant. This paper shows that the reactive power demand, estimated with an analytical model starting from the current/voltage waveforms foreseen in DEMO magnets, is huge. Therefore, an innovative approach, based on Active Front End converters, is proposed and compared with the traditional solution for coil power supplies. [ABSTRACT FROM AUTHOR]

RACE has been developing a Blanket Transporter concept for the replacement of tritium breeding blankets within the EUROfusion DEMO tokamak. This operation will be challenging due to the scale of the blankets (˜10 m & 80 tonnes) and the current vessel design, which necessitates complex blanket kinematics within the vessel prior to a high vertical lift through the upper port. The blanket transporter is a key high technical risk system and break-down rescue and recovery has been highlighted as a significant unaddressed risk. This paper outlines the processes used to identify the key hazards resulting from failure scenarios and the improvements made to the concept design to mitigate them. Failure modes were identified via industrial expert input, failure and hazard analyses. Mitigation of the failure scenarios was achieved by providing secondary load paths in the main structure, redundancy in key components and the inclusion of additional features that allowed secondary rescue equipment to engage to assist rescue. [ABSTRACT FROM AUTHOR]

In the framework of the EUROfusion Programme the EU is preparing the conceptual design of the fuel cycle for a pulsed fusion DEMO. Over the last years, a smart fuel cycle architecture has been developed, driven by the need to reduce the tritium inventory to an absolute minimum. To achieve this goal, batchwise processes used in the fusion fuel cycle so far were replaced by continuous processes wherever possible. This includes the change from discontinuous cryopumping to mercury based continuous vacuum pumping with zero demand on cryoplant power, and the introduction of temperature swing absorption processes for isotope separation in the tritium plant instead of large cryogenic distillation columns with tritiated liquid hold-ups. To circumvent handling of high inventories of tritium-bearing liquid fuel in cryogenic distillation columns of the tritium plant, superpermeable metal foils will be installed in the divertor ports of the vacuum vessel to extract pure DT continuously from the exhaust gas. Direct internal fuel recycling takes place via the pellet injection system. The paper discusses the design drivers and limitations, and describes the development status of the novel technologies. [ABSTRACT FROM AUTHOR]

• The progresses of the Primary Heat Transfer System of the HCPB Breeding Blanket for the DEMO reactor are outlined. • The main components of Ex-Vessel PHTS has been sized to be compliant with the new In-Vessel requirements. • Results highlights that the 2017 HCPB BB PHTS has reached improved thermal-hydraulic and safety characteristics. • Further efforts are still needed to reduce PHTS pressure drops and volumes as well as to improve IHXs performance. In the frame of the activities promoted and encouraged by the EURO-fusion Power Plant Physics and Technology (PPPT) department aimed at developing the EU-DEMO fusion reactor, strong emphasis has been recently posed 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, a very important role is played by the Breeding Blanket (BB) Primary Heat Transfer System (PHTS) as it is responsible to extract more than 80% of the fusion plasma power. In this framework, University of Palermo, Ansaldo Nucleare and CREATE have focused their work to improve thermal-hydraulic, safety and integration features of the Ex-Vessel PHTS for the Helium-Cooled Pebble Bed (HCPB) BB concept of DEMO. Starting from the outcomes obtained in 2016 that have allowed to highlight some criticalities which needed design changes, the paper describes progress and developments of the 2017 HCPB PHTS as well as the goals which have been achieved. The results of the research activity carried out show, in fact, i) an increase of the thermal-hydraulic performances, since a reduction in both total coolant inventory and total pressure drop has been respectively reached, ii) a potential improvement of the overall safety characteristics of the system. Nevertheless a critical assessment of these key parameters reveals that some issues are still open in terms of design integration and feasibility of the whole DEMO BoP for the helium-cooled blanket option, indicating that additional efforts are required to make this technology more attractive. [ABSTRACT FROM AUTHOR]

The divertor configuration defines the power exhaust capabilities of DEMO as one of the major key design parameters and sets a number of requirements on the tokamak layout, including port sizes, poloidal field coil positions, and size of toroidal field coils. It also requires a corresponding configuration of plasma-facing components (PFCs) and a remote handling scheme to be able to handle the cassettes and associated in-vessel components (IVC) the configuration requires. There is a risk that the baseline ITER-like single-null (SN) divertor configuration cannot meet the PFC technology limits regarding power exhaust and first wall protection while achieving the target plasma performance requirements of DEMO or a future fusion power plant. Alternative magnetic configurations – double-null, snowflake, X-, and super-X – exist and potentially offer solutions to these risks and a route to achievable power handling in DEMO. But these options impose significant changes on machine architecture, increase the machine complexity and affect remote handling and plasma physics and so an integrated approach must be taken to assessing the feasibility of these options. In this paper we describe the work programme to assess the requirements for incorporating these configurations into DEMO. [ABSTRACT FROM AUTHOR]

The paper presents the design activities and testing plan of a vertical target mock-up, developed within the pre-conceptual design phase for DEMO Work Package DIV-1 "Divertor Cassette Design and Integration" - EUROfusion Power Plant Physics & Technology (PPPT) program. Activities concerning the Divertor Outboard Vertical Target cooling mock-up are presented in term of CAD model, thermal-hydraulic numerical simulation, structural analysis, structural integrity verification and manufacturing procedure. Moreover, the mechanical dimensions of support systems for Plasma Facing Components (PFCs), manifold and diffuser have been analyzed in detail, in order to avoid structural fault during the test. Test procedures are discussed, taking into account design parameters, design code and facility performances. The CuCrZr alloy selected for the PFCs of EU DEMO divertor has been used also for the mock-up, while two options are still under evaluation for manifolds/diffuser, CuCrZr and stainless Steel 316 L(N)-IG, depending on the joining technology. Since the mock-up is mainly intended to verify hydraulic performances, it has been simplified by removing the W monoblocks from its PFCs. [ABSTRACT FROM AUTHOR]

The DEMO design has to face a number of challenges. According to the DEMO device limitations, the power load impinging on the Plasma Facing Components (PFCs) in the plasma Single Null (SN) magnetic configuration (which is for DEMO as in ITER the baseline configuration) clearly underlines the need to explore alternative divertor concepts. In this paper, a comparison between the Single and Double Null (DN) configurations is proposed. In particular, a comparative electromagnetic analysis of the magnetic configurations is performed in terms of vertical stability (VS) and disruptions in the 2D axisymmetric case. [ABSTRACT FROM AUTHOR]

The designs of DEMO components require the joining of low-activation steel (for example RUSFER EK-181, EUROFER, etc.) with tungsten. One of the most critical parameters is the thermal expansion mismatch between tungsten and steel, which can lead to the failure of components during use or even after manufacturing.In this paper, high-temperature brazing of EK-181 steel with tungsten was carried out. Pure copper was used for the direct joining of steel with tungsten. Brazing alloys based on copper (Cu–Ge and Cu–Ti) rapidly quenched into ribbon were used with a vanadium interlayer. Brazing was carried out in vacuum furnaces at 1100 °C for 20 min. The structural-phase states of the joints obtained were studied, the microhardness measured, shear strength tests carried out and thermocycling tests performed in the range 700–25 °C. [ABSTRACT FROM AUTHOR]

The vacuum vessel (VV) of DEMO has to fulfil various requirements. The main function is to provide the high vacuum environment essential for plasma operation. Likewise, it has important safety functions: it acts as the first and main nuclear confinement barrier and the main constituent of the neutron and radiation shielding function protecting other components in particular the superconducting coils and all maintenance areas. It has to remove the heat load from nuclear heating also in the unlikely event of a major technical malfunction. The VV is well designed to provide the mechanical support and attachment for the in-vessel components. This paper addresses the effect of the VV temperature on the functions and issues described, and the approach to define the best suitable temperature window. Following a qualitative assessment, the finding is that a lower operating temperature around 40 °C is advantageous for most of the issues. [ABSTRACT FROM AUTHOR]

Within the EUROfusion framework, a Molten Lead Ceramic Breeder (MLCB) breeding blanket is being developed at KIT as an alternative, cost-effective and low technology-risk solid breeder blanket for EU DEMO. The use of molten lead as neutron multiplier has been identified very promising, solving key potential issues coupled with the use of Be in solid breeder blankets. This MLCB concept has been derived from the enhanced Helium Cooled Pebble Bed, which is based on fission-like "fuel-breeder pin" configuration. Compared to former cooling-plate configuration, this concept greatly reduce the pressure drop, enabling the use of existing helium-circulator. After initial design studies with this concept for the previous EU DEMO baseline (BL2015), the status of the design activities on MLCB for the latest EU DEMO baseline (BL2017) is presented in this paper. First, neutronics analyses have been performed to assess and optimize the tritium breeding performance. Then, structural integrity under in-box LOCA has been assessed and the design has been iterated to fulfill the design criteria. Thereafter, thermo-hydraulic analyses have been conducted to evaluate the temperatures and pressure drops. After several design iterations to satisfy the temperature and pressure drop requirements, a structural assessment under normal condition has been conducted respecting the RCC-MRx rules. The results show that the current MLCB design fulfills the basic nuclear and thermo-mechanic-hydraulic requirements, setting the path for a consolidated design of this concept. Due to the absence of water-reactive elements in MLCB concept, water can be also used as coolant, which is under investigation. [ABSTRACT FROM AUTHOR]

The design study of DEMO is one of the main points of the European Roadmap to Fusion Electricity. The present pre-conceptual design phase of DEMO is used to explore a flexible range of the main machine geometrical design parameters and plasma scenarios. This paper presents the electromagnetic (EM) modelling of the DEMO baseline scenario, including the optimization activities on the electrically conductive structures surrounding the plasma and their influence on the passive vertical stabilization (VS). The improvements on the passive and active VS allowed increasing the maximum controllable plasma elongation, with a consequent increase on the fusion performance. Finally, a study is presented on the possibility to predict the plasma final position following a vertical displacement event (VDE). This prediction capability supports the development of a wall protection strategy from plasma transients. A model is presented based on the assumptions on the characteristic time constants expected for the current quench (CQ) time and L/R time constant of the conductive structures. [ABSTRACT FROM AUTHOR]

The EU-DEMO Water-Cooled Lithium Lead Breeding Blanket (WCLL BB) concept foresees liquid Pb-15.7Li eutectic alloy as breeder and neutron multiplier, whereas pressurized subcooled water as coolant, with operative conditions typical of the PWR fission reactors (temperature in the range of 295–328 °C and pressure of 15.5 MPa). The cooling down of the BB is guaranteed by means of two separated cooling circuits: the one consisted in square channels housed within the complex of Side Walls and First Wall, and the one composed of a set of Double-Walled Tubes (DWTs) submerged in the Breeding Zone (BZ) and deputed to remove heat power therein generated. A parametric thermal study has been carried out in order to optimize the DWTs lay-out within the BZ of the WCLL BB outboard equatorial region ensuring that EUROFER components maximum temperature stays below the allowable value of 550 °C. The analysis has been performed following a theoretical-numerical approach based on the Finite Element Method (FEM) and adopting the ABAQUS qualified commercial code. In the paper, results obtained together with the main assumptions adopted in the parametric analysis are critically discussed and a new, sustainable DWTs configuration is proposed. [ABSTRACT FROM AUTHOR]

To decrease the power density and associated high voltages, a distributed antenna system is proposed as ICRH system for the DEMO reactor. Among the different solutions, a layout made from a set of travelling wave array (TWA) sections is considered as the most promising. It optimizes coupling to the plasma, is load resilient and avoids large values for the VSWR in the feeding lines. The total radiated power scales as the number of independently fed sections such that high reliability can be expected. The TWA concept for ICRH is innovative and very different from the traditional IC antennas. A test on WEST would provide a proof of principle of the validity of the TWA approach together with a comparison with the existing WEST IC antennas. The chosen geometry of the TWA section is compatible with one unit of the complete set designed for a future reactor. The paper describes the progress made in the preparation of a test on WEST along with the extrapolation for a future reactor like DEMO. A comparative modeling with the present antennas is also discussed and a preliminary RAMI analysis is introduced showing the promising positive impact of the TWA design on the RAMI scores. [ABSTRACT FROM AUTHOR]

• New project at ENEA of a 250 GHz, 0.5 MW Cyclotron Auto Resonance Maser (CARM). • Design of a feasible TE 11 /TE 04 mode conversion for the RF Cold Test of CARM. • New solution with only a serpentine and a rippled wall mode converter. • Optimization achieved with four codes (CST-MS, HFSS and two in-house codes). • Efficiency of 95% for the Serpentine and 97% for the Rippled Wall Mode Converters. Electron Cyclotron Resonance Heating (ECRH) systems in future fusion devices, like the DEMO-nstration reactor, foresee an operational frequency in the range 230–280 GHz to match the plasma characteristics. Cyclotron Auto Resonance Masers (CARMs) could represent an alternative to gyrotrons as effective source of mm and sub-mm waves and the project of a 250 GHz, 0.5 MW CARM has been undertaken at ENEA. Mode converters from the TE 10 in WR3 rectangular waveguide to the operational mode (TE 53) are required to perform the RF cold test of Bragg reflectors and cavities. In this paper an improved and feasible structure for the TE 11 to TE 04 conversion, in oversized circular waveguides, using only two transitions with profile and radius optimized in order to obtain high efficiencies and sufficient bandwidth, has been designed and the simulation results are presented. The first transition is a TE 11 /TE 01 serpentine mode converter in circular waveguide (efficiency of 95%) with average radius of 1.48 mm and an appropriate number of geometrical periods. The second one, a TE 01 /TE 04 rippled wall mode converter with efficiency of 97%, consists of a single conversion section with an average radius of 3 mm and sixteen sinusoidal periods instead of three conversion sections (TE 01 /TE 02 , TE 02 /TE 03 , TE 03 /TE 04) in cascade, thus reducing the complexity and the length of the full transition chain. [ABSTRACT FROM AUTHOR]

The identification of structures, systems and components (SSCs) performing safety functions is of paramount importance for an EU DEMO development in all phases of the project: design, fabrication, commissioning, operation, maintenance, inspections and tests. Early identification of safety relevant SSCs contributes in the correct definition of systems requirements, preliminary layout and design integration. It may also be exploited as a supporting criterion for selection among the plant variants, e.g. alternative blanket concepts. Those SSCs assigned to a safety importance class (SIC) will receive adequate attention during the different developing phases. A graduation of safety important SSCs is adopted for DEMO on the basis of the IAEA Guide No. SSG-30, on SSCs classification. Three safety importance classes (SIC-1, SIC-2 and SIC-3) are assigned on the basis of: a) the consequences the loss of the safety function could lead to, b) the involvement of the SSCs in preventing, detecting or mitigating incident or accident and, c) the involvement of the SSCs in bringing and maintaining the plant in a safe state. In this paper, the provisional safety classification of the Primary Heat Transfer System (PHTS) related to the Helium Cooled Pebble Bed (HCPB) and Water Cooled Lithium Lead (WCLL) blankets of DEMO is presented. [ABSTRACT FROM AUTHOR]

In the case of fusion devices with long pulses or continuous operation, conventional magnetic sensors, such as inductive sensors, have to be supplemented by steady-state magnetic field sensors. Therefore, the Hall sensors are considered as a part of the DEMO magnetic diagnostics. The paper reviews the present status of the DEMO steady-state magnetic field sensors based on the ceramics-metal Hall sensors with a sensing layer made of antimony. The mission of the Hall sensors and their operating requirements are summarized, and the design of the Hall sensor assembly is presented. Satisfactory results of initial tests of the sensors are described. [ABSTRACT FROM AUTHOR]

In the present configuration of the European demonstration nuclear fusion power plant (DEMO), the toroidal magnetic field is produced by 16 superconducting toroidal field (TF) coils. The total stored energy of 146 GJ, more than 3 times the energy stored in ITER TF coils, is quickly dissipated by suitable Fast Discharge Units (FDU) whenever a coil fast discharge is required. The voltage applied to each coil during the discharge is defined by the energy, the current and the discharge time constant; the total circuit voltage is equally distributed among the coils thanks to the TF circuit symmetry; however, in case of fault the peak value applied to the coils and at the coil terminals to ground can be much higher. This paper deals with first studies addressed to the evaluation of the relative merit of different TF circuit topologies, based on the estimation of the maximum voltage stresses on the coils in various operating and fault conditions. Aim of the study is to support the coil design work, and to explore the best compromise between requirements for the coil insulation and cost and size of FDUs. Analyses and numerical simulations have been carried out, for both the cases of TF circuit composed of 16 and 8 sectors. The results are reported and discussed. [ABSTRACT FROM AUTHOR]

In the framework of the European "HORIZON 2020" program, EURO fusion develops a fusion power demonstrator (DEMO). According to a recent study about plasma heat loads, it appeared that the First Wall, which is the first part of the Breeding Blanket in front of the plasma, will face some high heat fluxes on some poloidal locations of the Breeding Blanket. In order to cope such high heat fluxes, this paper presents the investigation on different Helium cooled First Wall design integrated to the Breeding Blanket on the basis of standard square and circular smooth channels and with different options for the Tungsten armor surrounding the channels, taking advantages to the different material properties. The performances of the different concepts have been assessed with thermal and mechanical Finite Element Method numerical simulation based on a slice of the Helium Cooled Lithium Lead concept. Results are compared with the RCC-MRx design rules to prevent failure during normal and accidental condition. The results show that the options with channels surrounded by Tungsten could meet some plasma heat loads requirements from design point of view. However, the concept is still at an early stage of development and open issues are discussed. [ABSTRACT FROM AUTHOR]

• Estimation of the area and volume required to perform ex-vessel operations of transportation. • Comparison between the space required for nominal and rescue operations of transportation. • Preliminary estimation of the time duration for ex-vessel operations of transportation. The ex-vessel Remote Maintenance Systems are responsible for the replacement and transportation of the plasma facing components. The space required to perform the ex-vessel operations of transportation, as well as the respective time duration, has a great impact in the economics of a fusion power plant. This paper provides a preliminary assessment of the volume required to perform the operations of transportation along optimized trajectories in all levels of DEMO's reactor building. The overhead systems and ground systems are considered as the possible transportation technologies. In particular to the upper level, a cask concept is compared with a hot cell concept. Computed results of occupied volumes and time duration are presented taking into account the nominal, i.e., the expected and planned operations, and the possible rescue operations. [ABSTRACT FROM AUTHOR]

This paper presents the nuclear analysis carried out with the TRIPOLI-4® Monte-Carlo code for the Helium Cooled Lithium Lead (HCLL) "Advanced-Plus" breeding blanket design using the Single Module Segment (SMS) option in the European DEMO 2017 baseline. Compared to the previous one, this baseline is characterized by a radial outboard breeding blanket thickness reduction of 30 cm. Previous study has quantified its impact on Tritium Breeding Ratio (TBR reduction up to -0.08). This major constraint lead to the need of SMS solution development with HCLL "Advanced-Plus" design to reduce the amount of steel in the breeding blanket for TBR improvement. HCLL "Advanced-Plus" design is currently developed with the aim to improve the TBR. The main nuclear quantities: neutron wall loading, TBR, nuclear heating, neutron flux, displacement damage and helium production are reported and discussed. [ABSTRACT FROM AUTHOR]

• Pre-load structure of the DEMO pre-conceptual Central Solenoid design has been assessed. • ITER like materials has been considered for the tie-plates and conductor jacket. • Static Stress state of the CEA winding pack design proposal has been analyzed. The pre-conceptual design of the European DEMO fusion tokamak is currently being developed under the coordination of the EUROfusion Consortium. This paper reports the mechanical analysis of the Central Solenoid (CS), which comes right after the phase of definition of the winding pack proposed by the CEA. The first step consists in a draft-sizing of the tie-plates pre-compression structure. Then a global Finite element Analysis (FEA) of the CS, considering a sector-symmetry model and smeared winding pack properties is built-up in order to check this pre-load structure, and to identify the most critical loading scenario for the conductor. Finally, an axisymmetric local FEA of the CS winding pack is performed in order to study in detail the stress level in the conductor jacket and in the insulations layers. This pre-conceptual analysis procedure allowed to assess the acceptability of stress for the CEA winding pack proposal, and will be operated during the future magnetic system design iterations. [ABSTRACT FROM AUTHOR]

This paper analyses some existing industrial-scale processes for recycling of DEMO steel (Eurofer, AISI 316L) components and Lithium orthosilicates breeder, aiming at the evaluation of their practical feasibility and the individuation of preparatory activities to be performed. The thermodynamic analysis of recovering 14C from Eurofer and AISI 316L steels, by decarburization processes, was performed, based on practices used in steelmaking. The mass of species in the formed phases: gas, oxide, metal, and secondary wastes were quantified. The recovery of the expensive 6Li and the removal of impurities from used Li 4 SiO 4 solid breeder pebbles was analysed aiming at their reuse in the refabricating of fresh solid breeder. The study also aimed to identify and select existing and emerging remote handling technologies adopted in the modern steelmaking industry; in particular the refractory lining for casting moulding, maintenance repairing and replacing have been studied in detail. [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]