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Two concepts have been chosen to be tested in ITER under the form of Test Blanket Modules (TBMs): the Helium-Cooled Lithium-Lead (HCLL) and the Helium-Cooled Pebble-Bed (HCPB). Both European TBMs designs share similar steel box structure which is constituted by a box, made of two Side Caps (SCs) and a First Wall (FW), stiffened by horizontal and vertical Stiffening Plates (SP) and closed on its back by several back plates (BPs). All structure subcomponents are internally cooled by Helium circulating in meandering squared section channels. This paper describes manufacturing technologies developed and implemented to assembly the SPs into the box. It presents the preliminary manufacturing procedure developed and applied for the assembly of the SPs into the box by Tungsten Inert Gas (TIG). Several mock-ups have been manufactured from laboratory to feasibility mock-ups (scale 1:1) on which non-destructive and destructive tests have been carried-out to identify the preliminary manufacturing procedure. Due to TBM specificities (namely complex welding trajectories, heavy and big components, plates with channels, space constraints, …) a specific welding facility including a custom welding torch and an automated bench has been achieved and is also described in the paper. We detail the adopted manufacturing strategies, as the optimization of welding sequence to minimize distortions and the customization of welding parameters, to compensate machining tolerances and welding gaps. Results such as welded joints quality and microstructure, internal cooling channel deformation and structure distortions are reported. These developments have been performed following a standardized procedure complying with professional codes and standards (RCC-MRx).

The maximum heat load capacity of a DEMO First Wall (FW) of reasonable cost may impact the decision of the implementation of limiters in DEMO. An estimate of the engineering limit of the FW heat load capacity is an essential input for this decision. This paper describes the work performed to optimize the FW of the Water Cooled Lithium-Lead (WCLL) blanket concept for DEMO fusion reactor in order to increase its maximum heat load capacity. The optimization is based on the use of water at typical Pressurised Water Reactors conditions as coolant. The present WCLL FW with a waved plasma-faced surface and with circular channels was studied and the heat load limit has been predicted with FEM analysis equal to 1.0 MW m-2 with respect to the Eurofer temperature limit. An optimization study was then carried out for a flat FW design considering thermal and mechanical constraints assuming inlet and outlet temperatures equal to 285 °C/325 °C respectively and based on geometric design parameters such as channel pitch, diameter of pipes and thicknesses. It became clear through the optimization that the advantages of a waved FW are diminished. Given the manufacturing issues of that concept, the waved FW was therefore not pursued further. Even if the optimization study shows that the maximum heat load could in principle be as high as 2.53 MW m-2, it is reduced to 1.57 MW m-2 when additional constraints are introduced in order not to affect corrosion, manufacturability and Tritium Breeding Ratio in normal condition such as a coolant velocity ≤8 m/s, pipe diameter ≥5 mm and a total FW thickness ≤22 mm. However it is important to note that the FW channels currently fulfill additional functions and are therefore not optimized "at all cost" regarding heat load capacity and the paper points out some recommendations against missing assumptions. © 2015 Elsevier B.V. All rights reserved.

International audience; Two breeding blanket concepts, the Helium-Cooled Lithium-Lead (HCLL) and the Helium-Cooled Pebble-Bed (HCPB), developed in Europe in the frame of DEMO and relevant TBM studies, present many similarities in terms of design and manufacturing. As an example, all structure sub-components are internally cooled by helium circulating in meandering squared section channels. Several technologies have been investigated in the frame of EU fusion research programme for the manufacturing, including a specialized machining, fusion and diffusion welding and joints inspection, of DEMO breeding blanket and TBM sub-components (Cooling Plates (CP), Side Caps (SCs), Stiffening Plates (SPs), First Wall (FW)) and their assembly, main of which are discussed in this paper. The manufacturability of DEMO blanket modules Back Supporting Structure (BSS), a big structure located behind blanket modules and supporting them, is also discussed. The applicability of technologies takes into account specificities of EUROFER97 steel, foreseen as structural material.For each assessed technology, main results, e.g. in terms of mechanical properties and microstructure of weld joints, are presented and main advantages and drawbacks are summed up, in order to identify most promising technology/ies and to propose manufacturing scenarios. It should be noted that these developments are performed according to standards and professional codes (RCC-MRx). Further development strategies are also briefly discussed in the paper.

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.

To prepare the DEMO conceptual design phase a number of physics and engineering assessments were carried out in recent years in the frame of EFDA concluding in an initial design configuration of a DEMO tokamak. This paper gives an insight into the identified engineering requirements and constraints and describes their impact on the selection of the technologies and design principles of the main tokamak components. The EU DEMO program aims at making best use of the technologies developed for ITER (e.g., magnets, vessel, cryostat, and to some degree also the divertor). However, other systems in particular the breeding blanket require design solutions and advanced technologies that will only partially be tested in ITER. The main differences from ITER include the requirement to breed, to extract, to process and to recycle the tritium needed for plasma operation, the two orders of magnitude larger lifetime neutron fluence, the consequent radiation dose levels, which limit remote maintenance options, and the requirement to use low-activation steel for in-vessel components that also must operate at high temperature for efficient energy conversion. © 2015 Elsevier B.V. All rights reserved.

Within the framework of DEMO R&D activities, a research cooperation has been launched between ENEA, the University of Palermo and CEA to investigate the thermo-mechanical behaviour of the outboard equatorial module of the DEMO1 Water-Cooled Lithium Lead (WCLL) blanket under normal operation steady state scenario. The research campaign has been carried out following a theoretical-computational approach based on the Finite Element Method (FEM) and adopting a qualified commercial FEM code. In particular, two different 3D FEM models (Model 1 and Model 2), reproducing respectively the central and the lateral poloidal-radial slices of the WCLL blanket module, have been set up. A particular attention has been paid to the modelling of water flow domain, within both the segment box channels and the breeder zone tubes, to simulate realistically the coolant-box thermal coupling. Results obtained are herewith reported and critically discussed. © 2015 Elsevier B.V. All rights reserved.

SYCOMORE, a fusion reactor system code based on a modular approach is under development at CEA. Within this framework, this paper describes the relevant sub-modules which have been implemented to model the main outputs of the breeding blanket and shield block of the system code: tritium breeding ratio, peak energy deposition in toroidal field coils, reactor layout and power deposition, blanket pressure drops and materials inventory. Blanket and shield requirements are calculated by several sub-modules: the blanket assembly and layout sub-module, the neutronic sub-module, the blanket design sub-module (thermal hydraulic and thermo-mechanic pre-design tool). A power flow module has also been developed which is directly linked to the blanket thermo-dynamic performances, which is not described in this paper. For the blanket assembly and layout and the blanket module design sub-modules, explicit analytic models have been developed and implemented; for the neutronic sub-module neural networks that replicate the results of appropriate simplified 1D and 2D neutronic simulations have been built. Presently, relevant model for the Helium Cooled Lithium Lead is available. Sub-modules have been built in a way that they can run separately or coupled into the breeding blanket and shield module in order to be integrated in SYCOMORE. In the paper, the objective and main input/output parameters of each sub-module are reported and relevant models discussed. The application to previous studied reactor models (PPCS model AB, DEMO-HCLL 2006–2007 studies) is also presented.

The performed investigation focus on a monoblock type design for a water cooled DEMO divertor using Eurofer as structural material. In 2013, a study case of such a concept was presented. It was shown that basic concepts using Eurofer as structural material are limited to an incident heat flux of 8 MW m -2. Since, the EFDA agency issued new specifications. In this study, the conceptual design is reassessed with regard to specifications. Then, steady state thermal analyses and thermo-mechanical elastic analyses have been performed to define an upgrade of the geometry taking into account new specifications, design criteria and the maximum heat flux requirement of 10 MW m-2. An analysis of the influence of each adjustable geometrical parameter on thermo-mechanical design criteria was performed. As a consequence, geometrical parameters were set in order to fit to materials requirements. For defined hydraulic conditions taken in the most favourable configuration, the limit of this design is estimated to an incident heat flux of 10 MW m -2. Margin to critical heat flux and rules against progressive deformation/ratcheting in structural material limit the design. © 2014 Elsevier B.V.

The findings of the EU Fusion Programme's 'Materials Assessment Group' (MAG), assessing readiness of Structural, Plasma Facing (PF) and High Heat Flux (HHF) materials for DEMO, are discussed. These are incorporated into the EU Fusion Power Roadmap [1] , with a decision to construct DEMO in the early 2030s. The methodology uses project-based and systems-engineering approaches, the concept of Technology Readiness Levels, and considers lessons learned from Fission reactor material development. 'Baseline' materials are identified for each DEMO role, and the DEMO mission risks analysed from the known limitations, or unknown properties, associated with each baseline material. RD programmes to address these risks are developed. The DEMO assessed has a phase I with a 'starter blanket': the blanket must withstand >= 2 MW yr m(-2) fusion neutron flux (equivalent to similar to 20 dpa front-wall steel damage). The baseline materials all have significant associated risks, so development of 'Risk Mitigation Materials' (RMM) is recommended. The RD programme has parallel development of the baseline and RMM, up to 'down-selection' points to align with decisions on the DEMO blanket and divertor engineering definition. ITER licensing experience is used to refine the issues for materials nuclear testing, and arguments are developed to optimise scope of materials tests with fusion neutron ('14MeV') spectra before DEMO design finalisation. Some 14 MeV testing is still essential, and the Roadmap requires deployment of a >= 30 dpa (steels) testing capability by 2026. Programme optimisation by the pre-testing with fission neutrons on isotopically- or chemically-doped steels and with ion-beams is discussed along with the minimum 14 MeV testing programme, and the key role which fundamental and mission-oriented modelling can play in orienting the research. (C) 2013 Elsevier B.V. All rights reserved.

3D Monte Carlo transport codes are extensively used in neutronic analysis, especially in radiation protection and shielding analyses for fission and fusion reactors. TRIPOLI-4® is a Monte Carlo code developed by CEA. The aim of this paper is to show its capability to model a large-scale fusion reactor with complex neutron source and geometry. A benchmark between MCNP5 and TRIPOLI-4®, on the ITER A-lite model was carried out; neutron flux, nuclear heating in the blankets and tritium production rate in the European TBMs were evaluated and compared. The methodology to build the TRIPOLI-4® A-lite model is based on MCAM and the MCNP A-lite model. Simplified TBMs, from KIT, were integrated in the equatorial-port. A good agreement between MCNP and TRIPOLI-4® is shown; discrepancies are mainly included in the statistical error.

The water cooled lithium lead (WCLL) blanket, based on near-future technology requiring small extrapolation from present-day knowledge both on physical and technological aspect, is one of the breeding blanket concepts considered as possible candidates for the EU DEMOnstration power plant.In 2012, the EFDA agency issued new specifications for DEMO: this paper describes the work performed to adapt the WCLL blanket design to those specifications.Relatively small modules with straight surfaces are attached to a common Back Supporting Structure housing feeding pipes. Each module features reduced activation ferritic-martensitic steel as structural material, liquid Lithium-Lead as breeder, neutron multiplier and carrier. Water at typical Pressurized Water Reactors (PWR) conditions is chosen as coolant.A preliminary design of the equatorial outboard module has been achieved. Finite elements analyses have been carried out in order to assess the module thermal behavior. Two First Wall (FW) concepts have been proposed, one favoring the thermal efficiency, the other favoring the manufacturability. The Breeding Zone has been designed with C-shaped Double-Walled Tubes in order to minimize the Water/Pb-15.7Li interaction likelihood.The priorities for further development of the WCLL blanket concept are identified in the paper.

A methodology has been developed to consistently investigate, taking into account main reactor components, possible magnet solutions for a pulsed fusion reactor aiming at a large solenoid flux swing duration within the 2–3 h range. In a conceptual approach, investigations are carried out in the equatorial plane, taking into account the radial extension of the blanket-shielding zone, of the toroidal field magnet system inner leg and of the central solenoid for estimation of the pulsed swing. Design criteria are presented for the radial extension of the superconducting magnets, which is mostly driven by the structures (casings and conductor jacket). Typical available cable current densities are presented as a function of the magnetic field and of the temperature margin. The magnet design criteria have been integrated into SYCOMORE, a code for reactor modeling presently in development at CEA/IRFM in Cadarache, using the tools of the EFDA Integrated Tokamak Modeling task force. Possible solutions are investigated for a 2 GW fusion power reactor with different aspect ratios. The final adjustment of the DEMO pulsed reactor parameters will have to be consistently done, considering all reactor components, when the final goals of the machine will be completely clarified.

SYCOMORE, a fusion reactor system code based on a modular approach, is under development at CEA. In this framework, this paper describes a methodology developed to build the neutronic module of SYCOMORE. This neutronic module aims to evaluate main neutronic parameters characterising a fusion reactor (tokamak): tritium breeding ratio, multiplication factor, nuclear heating as a function of the reactor main geometrical parameters (major radius, elongation, etc.), of the radial build, Li enrichment, blanket and shield thickness, etc. It is based on calculations carried out with APOLLO2 and TRIPOLI-4 CEA transport code on simplified 1D and 2D neutronic models. These models are validated versus a more detailed 3D Monte-Carlo model (using TRIPOLI-4). To ease the integration of this neutronic module in SYCOMORE and provide results instantly, a surrogate model that replicates the 1D and 2D neutronic model results was used. Among the different surrogate models types (polynomial interpolation, responses functions, interpolating by Kriging, artificial neural network, etc.) the neural networks were selected for their efficiency and flexibility. The methodology described in this paper to build SYCOMORE neutronic module is devoted to HCLL blanket, but it could be applied to any breeder blanket concept provided that appropriate validation could be carried out.

In this paper water-cooled divertor concepts based on tungsten monoblock design identified in previous studies as candidate for fusion power plant have been reviewed to assess their potential and limits as possible candidates for a DEMO concept deliverable in a short to medium term ("conservative baseline design"). The rationale and technology development assumptions that have led to their selection are revisited taking into account present factual information on reactor parameters, materials properties and manufacturing technologies. For that purpose, main parameters impacting the divertor design are identified and their relevance discussed. The state of the art knowledge on materials and relevant manufacturing techniques is reviewed. Particular attention is paid to material properties change after irradiation; phenomenon thresholds (if any) and possible operating ranges are identified (in terms of temperature and damage dose). The suitability of various proposed heat sink/structural and sacrificial layer materials, as proposed in the past, are re-assessed (e.g. with regard to the possibility of reducing peak heat flux and/or neutron radiation damages). As a result, potential and limits of various proposed concepts are highlighted, ranges in which they could operate (if any) defined and possible improvements are proposed. Identified missing point in materials database and/or manufacturing techniques knowledge that should be uppermost investigated in future R&D activities are reported. This work has been carried out in the frame of EFDA PPPT Work Programme activities. © 2013 Elsevier B.V.

Within the European Fusion Technology Program, the Water-Cooled Lithium Lead (WCLL) DEMO breeding blanket line was selected in 1995 as one of the two EU lines to be developed in the next decade, in particular with the aim of manufacturing a Test Blanket Module (TBM) to be implemented in ITER. This specific goal has been maintained also in ITER-FEAT program even if the general design parameters of the TBMs have reported some changes. This paper is focused on the investigation of the WCLL-TBM nuclear response in ITER-FEAT through detailed 3D-Monte Carlo neutronic and photonic analyses. A 3D heterogeneous model of the most recent design of the WCLL-TBM has been set-up simulating realistically its new lay out and taking into account 9% Cr martensitic steel as structural material. It has been inserted into an existing 3D semi-heterogeneous ITER-FEAT model accounting for a proper D–T neutron source. The analyses have been performed by means of MCNP-4C code running on a cluster of four workstations through the implementation of a parallel virtual machine. The main WCLL-TBM nuclear responses have been determined focusing the attention on power deposition density, material damage through displacement per atom (DPA) and He and H production rate, daily tritium production and tritium production rate radial distribution in the module. Moreover, the impact of using lithium at various Li6 enrichment on the TBM nuclear response has been investigated. The results obtained are herewith presented and critically discussed.