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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]

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]

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]

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]

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]

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 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]

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]

The paper describes the organization of the Test Blanket Module (TBM) program, its overall objective and schedule and the status of the technical activities within the ITER Organization-Central Team (IO-CT). The latter include the design integration of the Test Blanket Systems (TBSs) into the nuclear buildings, ensuring all interfaces with other ITER systems, the design of the common TBS components such as the TBM Frames, the Dummy TBMs, and the TBS maintenance tools and equipment in the TBM Port Cell as well as in the Hot Cell building, the design of the TBS connection pipes and the definition of the required maintenance operations and associated R&D. The paper also discusses the major challenges that the TBM Program will be facing in ITER such as the potential impact of the TBMs ferritic/martensitic structures on plasma operations, the approaches to tritium and contamination confinement, the required mitigation and recovery actions in case of accidents, and the assessment of the reliability aspects that could have an impact on ITER availability. [ABSTRACT FROM AUTHOR]

Abstract The current breeding blankets proposed in the different conceptual fusion power plants are based mainly on the use of Li 4 SiO 4 and/or Li 2 TiO 3 as tritium breeder and Be/Be 12 Ti as neutron multiplier or an eutectic Li 17 Pb 83 for as a hybrid tritium and neutron multiplier. While these materials offer some tritium breeding capabilities, some recent studies show that the tritium self-sufficiency may not be ensured with these materials due to the strong reduction of blanket coverage after the integration of other in-vessel reactor systems (heating and current drive, limiters, large or double-null divertor systems, etc.). Also, some materials like Be raises several key feasibility concerns. The goal of this paper is to perform an update of the screening for tritium breeder and neutron multiplier materials and to assess the tritium breeding performance of the selected compounds in order to reveal new options. As for the neutron multiplier materials, a new subdivision between solid and liquid multipliers is proposed. For the selected compounds, detailed 3D heterogeneous neutronic analyses have been performed with MCNP5-1.60 assuming the architecture of the current EU DEMO Helium Cooled Pebble Bed (HCPB) as a benchmark breeding blanket. From the point of view of ceramic breeders, Li 8 ZrO 6 has been found to outperform Li 4 SiO 4 by more than 4% in terms of tritium breeding, having 6% higher melting point. From the point of view of solid neutron multipliers, Be 12 Cr, Be 12 V, Be 13 Zr and Be 13 Y show a similar performance as Be 12 Ti, while LaPb 3 , Zr 5 Pb 4 and YPb 2 offer a solution for a Be-free blanket. As for liquid multipliers, Pb in combination with a ceramic breeder shows a very promising option. Moreover, Pb compounds like Pb 90 Mn 10 and Pb 95 Ba 5 offer similar performance as Pb with a lower melting point (290 °C). Due to the significant advantages of molten Pb as neutron multiplier, future work will be conducted to define a design of a helium cooled Molten Lead Ceramic Breeder blanket, as simple, cost effective blanket concept. [ABSTRACT FROM AUTHOR]

Highlights • Description of the manifolds of the HCLL BB is presented. • CFD analysis is performed to optimize the geometries of the manifolds of HCLL BB. • Effect of the tie rods, number of outlets, and number of inlets is found in the manifolds. • The final optimized geometry, reduced the pressure drop by ∼ 60% in both MF1 and MF3. Abstract In Europe the technology to develop DEMO is being supported by HORIZON 2020 research programme under EUROfusion Consortium. One of the key components that DEMO relies on is a working Breeding Blanket (BB). There are currently four Breeding Blanket concepts being developed in Europe, whose primary purpose is to convert kinetic energy of the neutrons into heat and to transfer this converted heat to a coolant, breed tritium to make plant self-sufficient of fuel and to shield magnets from radiation. This paper describes helium distribution system in the "Optimized Conservative" concept of Helium Cooled Lithium Lead (HCLL) BB, according to the HCLL 2014 Design Description Document. This contains three manifolds to distribute the helium in to several parts of the BB namely the First Wall, Cooling Plates, horizontal Stiffening Plates and the vertical Stiffening Plates. The work mainly focuses on optimization study and reduction of the pressure drops in the helium manifolds. The same is achieved by changing several geometrical parameters to keep the pressure drops under the limit. To achieve the results ANSYS-FLUENT commercial Computational fluid dynamic code is used. The final results include the modified geometry which gives lowest pressure drops for the given mass flow rate. [ABSTRACT FROM AUTHOR]

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

Abstract Korea has own programs toward DEMO and fusion reactors that will require further improved DEMO blanket and energy utilization systems. Primary goals of DEMO blanket development in Korea are to develop and verify the integrated blanket design tools; to develop blanket materials, cooling and tritium fuel cycle technologies; and to develop and evaluate fabrication and joining technologies. The concept of helium-cooled ceramic reflector (HCCR) blanket is adopted to be tested in ITER as Test Blanket Module (TBM). Currently, the design and R&D activities are mainly performed through the ITER TBM program in Korea. It is expected to demonstrate the major objectives of the breeding blanket: extraction of heat from burning plasma, tritium self-sufficiency and its integrity within the whole systems considering safety features. Although TBM program will provide very essential data and experience, there will be a considerable technical gap from DEMO blanket. This paper presents the technical gap in the main technical categories of the breeding blanket based on the experience of the development of HCCR TBM and breeding blanket. It is found that still about 60–70% of the DEMO blanket technology can be achieved through the HCCR TBM depending on the technology maturity level, and some ways to DEMO blanket from HCCR TBM are proposed. [ABSTRACT FROM AUTHOR]

A global thermal-hydraulic model of the EU DEMO tokamak, based on an object-oriented approach using the Modelica language, is currently under development at Politecnico di Torino. The first module of the global model will simulate the dynamics of the blanket cooling system and it will be able to investigate different coolant options and different cooling schemes, adapted to the different blanket systems currently under development in the Breeding Blanket project. Following the development of the module related to the Helium-Cooled Pebble Bed blanket, this paper presents the dynamic model of the Water-Cooled Lithium Lead (WCLL) blanket, designed at ENEA Brasimone, together with the first results obtained in steady-state conditions. [ABSTRACT FROM AUTHOR]

ENEA CR Brasimone has developed the new design of the Water Cooled Lithium Lead Breeding Blanket (WCLL BB). In the new design Breeding Zone (BZ) water coolant flows in radial-toroidal direction, and PbLi flows in radial-poloidal direction; a gap between the Back Plate (BP) and the BZ constitutes the PbLi inlet manifold. The paper presents the CFD analysis of the WCLL BB PbLi inlet manifold, performed by ANSYS CFX-15 code. The objective of the analysis is to investigate the PbLi flow paths in the manifold region and to optimize the mass flow rate distribution in the BZ of the module. A preliminary analytical analysis is performed to evaluate pressure drop at orifices and to select the diameter of the orifices. CFD simulations are performed with the selected geometries. The optimal geometric configuration is a compromise between the need to have low PbLi velocity, to limit MHD issues and Eurofer corrosion, and to preserve the structural capability of the stiffening plates to withstand the overpressure conditions. Results demonstrate that the layout of the manifold and the size of the orifices is suitable to uniformly distribute the PbLi in the BZ, fulfilling the fluid velocity constraints. Specific MHD analyses are required to evaluate more realistic flow paths and to demonstrate the feasibility of the design. [ABSTRACT FROM AUTHOR]

The Helium Cooled Pebble Bed (HCPB) Breeding Blanket (BB) is one of the 4 BB concepts being investigated in the EU for their possible implementation in DEMO. During 2014 the former “beer-box” BB concept based on the ITER’s HCPB Test Blanket Module suffered several design changes so as to meet the different counteracting nuclear, thermohydraulic and thermomechanical requirements. These studies evidenced that the concept is too rigid to meet the tight TBR requirements imposed for the EU DEMO (i.e. TBR ≥ 1.10). Additionally, the complex manifold system with unbalanced helium mass flow in each of the 2 parallel cooling loops made the concept thermohydraulically complex. However, parametric studies during 2015 revealed that the HCPB concept have potential for a better nuclear performance, as well as margin for a significant simplification of the cooling internals by redefining the cooling plates and the architecture of the blanket, building a symmetric flow scheme. This paper describes the new HCPB concept based on an integrated FW with the breeding zone thermohydraulics and helium manifold systems. The former complex manifold backplates have been compacted and integrated in the cooling plates, releasing ≈300 mm of radial space that can be used now to increase breeder zone, the neutron shielding, to reinforce the Back Supporting Structure (BSS) or basically to reduce the reactor size. Detailed neutronic analyses have yielded a TBR of ∼1.20 for the baseline design. Initial analyses show a correct thermohydraulic behavior. Preliminary thermomechanical analyses also indicate that the design can potentially withstand an in-box LOCA at 9 MPa at a level C according to the RCC-MRx code. Future consolidation activities are described, which shall lead to a concept meeting the BB requirements. [ABSTRACT FROM AUTHOR]

In the frame of the EUROfusion Consortium activities were launched in 2014 to develop a concept of a DEMO reactor including a large R&D program and the integrated design of the tokamak systems. The integration of the in-vessel components (IVCs) must accommodate numerous constraints imposed by their operating environment, the requirements for precise alignment, high performance, reliability, and remote maintainability. This makes the development of any feasible design a major challenge. Although DEMO is defined to be a one-of-a-kind device there needs to be in addition to the development of the IVC design solutions a remarkable emphasis on the optimization of these solutions already at the conceptual level. Their design has a significant impact on the machine layout, complexity, and performance. This paper identifies design and technology limitations of IVCs, their consequences on the integration principles, and introduces strategies currently considered in the DEMO tokamak design approach. [ABSTRACT FROM AUTHOR]

The Lead-Lithium Ceramic Breeder Test Blanket Module (LLCB-TBM) is the Indian TBM representing the DEMO breeding blanket, to be installed in ITER radial port no-2 for testing in ITER. The conceptual design of LLCB TBM systems and their engineering design are under progress. The neutronic analysis and engineering design of LLCB TBM set (TBM + TBM Shield) is undergoing optimization. The LLCB TBS auxiliary systems; Helium cooling systems, Lead-Lithium Cooling System and Tritium Extraction Systems design are conceptualized and detail design is under progress. The system arrangements in port cell area, Tokamak Cooling Water System (TCWS) vault and Tritium building along with interface requirements have been worked out within the allocated space. LLCB TBS related R&D activities are under progress at Institute for Plasma Research (IPR), Gandhinagar in association with Bhabha Atomic Research Centre (BARC), Mumbai and Indira Gandhi Centre for Atomic Research, Kalpakkam. The major research and development areas are liquidmetal technologies, lithium ceramic pebbles, lead-lithium eutectic alloy, India specific Reduced Activation Ferritic Martensitic Steels (IN-RAFMS) development and fabrication technologies development. This paper will highlight the current LLCB TBM set and axillary systems design and status of R&D activities in various areas. [ABSTRACT FROM AUTHOR]

The conceptual design of the tritium extraction system (TES) for the European DEMO reactor is worked out in parallel for four different breeding blankets (BB) retained by EUROfusion. The TES design has to be tackled in an integrated manner optimizing the synergy with the directly interfacing inner fuel cycle, while minimizing the tritium permeation into the coolant. Considering DEMO requirements, it is most likely that only advanced technologies will be suitable for the tritium extraction systems of the BB. This paper overviews the European work programme for R&D on tritium technology for the DEMO BB, summaries the general first outcomes, and details the specific and comprehensive R&D program to study experimentally immature but promising technologies such as vacuum sieve tray or permeator against vacuum for tritium extraction from PbLi, and advanced inorganic membranes and catalytic membrane reactor for tritium extraction from He. These techniques are simple, fully continuous, likely compact with contained energy consumption. Several European Laboratories are joining their efforts to deploy several new experimental setups to accommodate the tests campaigns that will cover small scale experiments with tritium and inactive medium scale tests so as to improve the technology readiness level of these advanced processes. [ABSTRACT FROM AUTHOR]

The EUROfusion Consortium develops a design of a fusion power demonstrator (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 surrounding the plasma, ensuring tritium self-sufficiency, heat removal for conversion into electricity, and neutron shielding. The Helium Cooled Lithium Lead (HCLL) blanket is one of the concepts which is investigated for DEMO. It is made of a Eurofer structure and uses the eutectic liquid lithium–lead as tritium breeder and neutron multiplier, and helium gas as coolant. Within the EUROfusion organization, CEA with the support of Wigner-RCP and IPP-CR, is in charge of the design of the HCLL blanket for DEMO. This paper presents the status of the thermal and mechanical analyses carried out on the HCLL breeding blanket in order to justify the design. CFD thermal analyses on generic breeding unit including stiffening plates and cooling plates have been performed with ANSYS in order to consolidate results obtained with previous FEM design analyses. Moreover in order to expand the justification of the HCLL Breeding blanket design, the most loaded area of the equatorial outboard module has been analyzed with CAST3M in case of accident, according to RCC-MRx, resulting in relatively good behavior of the box. Additionally analyses have been performed in order to check the good behavior of the tie rods modules’ attachments system in normal condition showing promising results. [ABSTRACT FROM AUTHOR]

This paper describes the main acceptance criteria and required acceptance tests for the components of the six Test Blanket Systems to be installed and operated in ITER. It summarizes the guide-line toward the establishment of detailed test plans for the TBS, starting from the end-product at the ITER Members factories, and to generally define the type of tests that have to be performed on the ITER site after shipment, and/or prior to the systems final commissioning phase. [ABSTRACT FROM AUTHOR]

• The paper focuses on detailed assessment about design of the inboard wall of DEMO Vacuum Vessel. • The work evaluates both P and S type damages for the inboard wall of DEMO Vacuum Vessel. • The structural integrity of DEMO Vacuum Vessel inboard wall design has been checked against the assumed load conditions. • The elastic analysis method has been used to check the rules for prevention of both types of damage. • Optimization structural analyses were performed and checked against the rules provided in the applicable design code. The pre-concept design of the DEMO Vacuum Vessel is going on in view of the 2020 gate review; moreover the nuclear heat loads on the vessel inner shell were determined and found to be about one order of magnitude higher compared to ITER. A subsequent thermal-structural analysis of the vessel inner shell revealed high thermal stresses and a large temperature gradient through the inner shell thickness. Given the simultaneous occurrence of primary membrane stresses in the entire vessel inboard wall and, in proximity of the vessel ribs, high bending stresses due to the coolant pressure, a survey of all options within the design rules was required to identify the inter-dependencies of the individual stress limits (primary membrane, primary bending, thermal membrane plus bending). In order to face this kind of issues a detailed assessment on the design of the inboard wall of DEMO Vacuum Vessel has been conducted and is presented here. The current work evaluates both P and S type damages for the inboard wall of DEMO Vacuum Vessel in case of high nuclear heat load, vacuum vessel coolant pressure and toroidal field coil fast discharge. The elastic analysis method has been used to check the rules for prevention of both types of damage. The rules applied to prevent the aforementioned damages are compliant to Level A criteria, in case of negligible creep and negligible irradiation. In order to check the structural integrity of the inboard wall of DEMO VV against high thermal and mechanical loads, optimization structural analyses were performed and checked against the rules provided in the applicable design code (RCC MRx). [ABSTRACT FROM AUTHOR]

• The ITER TBM Program plans to test DEMO-relevant tritium breeding blanket systems. • The selection of the 4 TBSs for Initial Configuration is expected in a near future. • Each TBS is formed by various sub-systems with different functions and technologies. • The selection of the four InCo TBSs is based on the existing TBS conceptual designs. • The infrastructures hosting the TBSs has reached the preliminary design stage. The ITER Test Blanket Module (TBM) Program has significantly evolved since 2018. The number of equatorial ports allocated for operating the Test Blanket Systems (TBSs) has been reduced from three to two. As consequence, four TBSs can be simultaneously installed and operated, versus six previously. Since the dedicated space in the various rooms of the Tokamak Complex has been kept unchanged, the existing space constraints have been substantially relaxed. The paper addresses the possible selection of the four TBSs for the initial configuration and design of the involved TBMs and the main on-going R&D carried out by the ITER Members (IMs) in support of the TBS designs. It describes also the main progress of the design of the infrastructures needed for hosting the four TBSs (e.g., port plugs, port cell common components, common maintenance tools and equipment). The design of the TBS Connection Pipes (CPs) System has reached its final phase since, being captive, it needs to be installed before the First Plasma. [ABSTRACT FROM AUTHOR]

• SYCOMORE: a modular system code for fusion reactor. • The design module for a WCLL Breeding Blanket is implemented in SYCOMORE. • The design module in SYCOMORE will be a useful pre-conceptual design tool for designers. The pre-conceptual design of the DEMOnstration reactors has already started and several tokamak configurations have to be tested to find the best design by exploring different design parameters. Fast simulations involving the different components behavior must be performed. Within the European framework, SYCOMORE (SYstem COde for MOdelling tokamak REactor) is developed by CEA for this purpose. The Breeding Blanket (BB) facing the plasma is a key component in DEMO ensuring tritium self-sufficiency, shielding against neutrons and heat extraction for electricity production. Several BB concepts are being studied, among which the Water Cooled Lithium Lead (WCLL) one. SYCOMORE includes several specific modules in Python linked together, one of which has been developed to define a suitable design of the WCLL Breeding Blanket and is presented in this paper. The method to define automatically the WCLL First Wall (FW) design using analytical design formulae starting from thermo-hydraulic and thermo-mechanical considerations as well as design criteria coming from Codes & Standards (C&S) is presented. Furthermore WCLL FW design obtained with SYCOMORE is compared to Finite Elements (FE) analyses of the DEMO WCLL BB. Finally, a coupling between thermo-mechanics and neutronics is implemented, several iterations are necessary to obtain a converged design. Neutronic block evaluates the radial build, the BB tritium production, and the nuclear heating in the FW and the Breeding Zone (used by thermo-mechanical block). Thermo-mechanical module gives the design data (FW thickness, compositions, etc.) to the neutronic block. [ABSTRACT FROM AUTHOR]

The realization of a DEMOnstration Fusion Power Reactor (DEMO) to follow ITER, with the capability of generating several hundred MW of net electricity and operating with a closed fuel-cycle by 2050, is viewed by Europe as the remaining crucial step towards the exploitation of fusion power. The EUROfusion Consortium, in the frame of the European Horizon 2020 Program, has been assessing four different breeding blanket concepts in view of selecting the reference one for DEMO. This paper describes technologies and manufacturing scenarios developed and envisaged for the four blanket concepts, including nuclear "conventional" assembly processes as GTAW, electron beam and laser welding, Hot Isostatic Pressing (HIP), and also more advanced (from the nuclear standpoint) technologies as additive manufacturing techniques. These developments are performed in conformity with international standards and/or design/manufacturing codes. Topics as the metallurgical weldability of EUROFER steel and the associated risks or the development of appropriate filler wire are discussed. The development of protective W-coating layers on First Wall, with Functionally Graded (FG) interlayer as compliance layer between W and EUROFER substrate, realized by Vacuum Plasma Spraying method, is also propounded. First layer systems showed promising layer adhesion, thermal fatigue and thermal shock properties. He-cooled mock-ups, representative of the First Wall with FG W/EUROFER coating are developed for test campaigns in the HELOKA facility under relevant heat fluxes. First elements of Double Walled Tubes (DWT) manufacturing and tube/plate assembly for the water cooled concept are given, comprising test campaign aiming at assessing their behaviour under corrosion. In addition, further development strategies are suggested. [ABSTRACT FROM AUTHOR]