Search

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

A major part in the EUROfusion materials research program is dedicated to characterize and quantify nuclear fusion specific neutron damage in structural materials. While the majority of irradiation data gives a relatively clear view on the displacement damage, the effect of transmutation – i.e. especially hydrogen and helium production in steels – is not yet explored very well. However, few available results indicate that EUROFER-type steels will reach their operating limit as soon as the formation of helium bubbles reaches a critical amount or size. At that point, the material would fail due to embrittlement at the considered load. This paper presents a strategy for the mitigation of the before-mentioned problem using the following facts: • the neutron dose and related transmutation rate decreases quickly inside the first wall, that is, only a plasma-near area is extremely loaded • nanostructured oxide dispersion strengthened (ODS) steels may have an enormous trapping effect on helium and hydrogen, which would suppress the formation of large helium bubbles • compared to conventional steels, ODS steels show improved irradiation tensile ductility and creep strength In summary, producing the plasma facing, highly neutron and heat loaded part of blankets by an ODS steel, while using EUROFER97 for everything else, would allow a higher heat flux as well as a longer operating period. Consequently, we (1) developed and produced 14 % Cr ferritic ODS steel plates. (2) We fabricated a mockup with 5 cooling channels and a plated first wall of ODS steel, using the same production processes as for a real component. And finally, (3) we performed high heat flux tests in the HELOKA facility (Helium Loop Karlsruhe at KIT) applying short and up to 2 h long pulses, in which the operating temperature limit for EUROFER97 (i.e., 550 °C) was finally exceeded by 100 K. Thereafter, microstructure and defect analyses did not reveal defects or recognizable damage. Only a heat affected zone in the EUROFER/ODS steel interface could be detected. This demonstrates that the use of ODS steel could make a decisive difference in the future design and performance of breeding blankets.

In future nuclear fusion reactors, plasma facing components have to sustain specific neutron damage. While the majority of irradiation data provides a relatively clear picture of the displacement damage, the effect of helium transmutation is not yet explored in detail. Nevertheless, available results from simulation experiments indicate that 9%-chromium steels will reach their operating limit as soon as the growing helium bubbles extent a critical size. At that point, the material would most probably fail due to grain boundary embrittlement. In this contribution, we present a strategy for the mitigation of the before-mentioned problem using the following facts. (1) The neutron dose and related transmutation rate decreases quickly inside the first wall of the breeding blankets, that is, only a plasma-near area is extremely loaded. (2) Nanostructured oxide dispersion strengthened (ODS) steels may have an enormous trapping effect on helium, which would suppress the formation of large helium bubbles for a much longer period. (3) Compared to conventional steels, ODS steels also provide improved irradiation tensile ductility and creep strength. Therefore, a design, based on the fabrication of the plasma facing and highly neutron and heat loaded parts of blankets by an ODS steel, while using EUROFER97 for everything else, would extend the operating time and enable a higher heat flux. Consequently, we (i) developed and produced 14%Cr ferritic ODS steel plates and (ii) optimized and demonstrated a scalable industrial production route. (iii) We fabricated a mock-up with five cooling channels and a plated first wall of ODS steel, using the same production processes as for a real component. (iv) Finally, we performed high heat flux tests in the Helium Loop Karlsruhe, applying a few hundred short and a few 2 h long pulses, in which the operating temperature limit for EUROFER97 (i.e. 550 ◦C) was finally exceeded by 100 K. (v) Thereafter, microstructure and defect analyses did not reveal critical defects or recognizable damage. Only a heat affected zone in the EUROFER/ODS steel interface could be detected. However, a solution to prohibit the formation of such heat affected zones is given. These research contributions demonstrate that the use of ODS steel is not only feasible and affordable but could make a decisive difference in the future design and performance of breeding blankets.

International audience; In the European strategy, DEMO is the intermediate step between ITER and a commercial fusion power plant. In this framework, one of the goal of DEMO is to be a Breeding Blanket test facility. The Breeding Blanket, which is not present in ITER, is one of the key components for the future deployment of nuclear fusion electricity as it accomplishes the functions of tritium breeding and nuclear to thermal power conversion.Due to time constraints lead by the construction schedule of DEMO, a new strategy to consider in DEMO a “driver” Breeding Blanket that needs limited technological extrapolation has been chosen, while “advanced” Breeding Blanket concepts will be tested in the next phases. In this context, ITER will allow to test technologies to provide relevant contributions in terms of Return of eXperience to the DEMO “driver” Breeding Blanket project by the mean of Test Blanket Modules (TBM) to be installed in different ITER Vacuum Vessel Ports.Among the possible “driver” Breeding Blanket, the Water Cooled Lithium Lead (WCLL) concept comes out. In this framework, a realignment of the DEMO Breeding Blanket and TBM programs has started in 2017, leading to a new TBM development relevant of the DEMO WCLL BB. The WCLL TBM is therefore an essential component in ITER that will provide crucial information for the development of the DEMO “driver” blanket.This paper aims at presenting the development process and design status of WCLL TBM. After recalling the main features of the WCLL TBM, conceptual design analyses are presented and discussed.

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.

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).

Thermohydraulic measurements in liquid metal test blanket modules for ITER require reliable sensors for detection of temperature and magnetohydrodynamic properties of the flow such as the distribution of electric potential. The latter one can be directly compared with numerical simulations or interpreted in terms of local liquid metal flow rates. Combined temperature-potential probes may serve for both purposes for sensing temperature and potential values, while remaining relatively small in size. By comparison with results obtained using other sensing techniques or theoretical results, the present paper confirms that such measurements in magnetohydrodynamic liquid metal flows may yield accurate data for potential.

In the framework of the European “HORIZON 2020” research program, the EUROfusion Consortium develops a design of a fusion demonstrator (DEMO). The Breeding Blanket (BB) directly surrounding the plasma is a major component ensuring tritium self-sufficiency, shielding against neutrons from D-T plasmas and heat extraction for electricity conversion. CEA-Saclay, with the support of Wigner-RCP and IPP-CR is responsible of developing the Helium Cooled Lithium Lead (HCLL) BB concept. In order to enhance Tritium Breeding Ratio (TBR), an alternative HCLL BB design, called “Advanced-Plus” concept, has been investigated, with the aim to reduce the amount of steel, while ensuring good thermal and mechanical behavior. The thermal behaviors of the horizontal Stiffening Plates (hSPs) with various channels designs in terms of dimensions and layout have been parsed on simplified FE models for verifying and optimizing the HCLL Breeding Zone (BZ) before analyzing a HCLL BB ¼ model. Moreover, sensitive analyses have been carried out for testing various hSPs pitches, alternative Helium hydraulic schemes inside the hSPs as well as the detachable FW option in applying a zero Heat Flux (HF) on the FW. Furthermore, mechanical calculations have also been performed to assess the behavior of the module in faulted condition. The results obtained with the Cast3M-qualified-FEM code are presented and discussed.

International audience; The next step for fusion energy after the ITER tokamak is the demonstration power plant DEMO. In this framework , system codes are used to address high-level key design issues for the DEMO pre-conceptual phase. They aim at capturing the interactions between the subsystems of a fusion reactor. SYCOMORE is a modular system code which includes physics and technology models coupled to an optimizer in order to explore a large design parameter space. In the present paper, trade-off studies focused on technology modules are reported including the influence of some design-driving assumptions on the reactor performances and size, starting from a European DEMO1-like design (more than 500 MW net electric power and 2 h burn duration). The increase of the mechanical stress limits in TF and CS magnets can help reducing the reactor size, slightly more when high temperature superconductors are used in the TF coil. The tritium breeding ratio can be improved to more than 1.10 by a moderate increase of the size, but the tritium burn-up ratio needs one additional meter of major radius for every percent increase. Divertor coolant options are also compared, showing some differences between helium, hot and cold water scenarios at various incident divertor heat fluxes.

International audience; The next step after ITER is the demonstration of stable electricity production with a fusion reactor. Key design performances will have to be met by the corresponding power plant demonstrator (DEMO), fulfilling a large number of constraints. System codes such as SYCOMORE, by simulating all the fusion power plant subsystems , address those questions. To be able to perform design optimizations, simplified models relying on physical and technological assumptions have to be used, resulting in a large number of input parameters. As these parameters are not always exactly known, the impact of their associated uncertainties on final design performances has to be evaluated. Sensitivity methods, by measuring the relative influence of inputs on the figures of merit of the design, allow to select the dominant parameters. This information helps the search for optimal working points, guides the priority for technical improvements and finally allows selecting meaningful inputs for uncertainty propagation. A full set of sensitivity methods and their application on a ITER and a DEMO design will be presented, discussing both the statistical methods behaviors and the physical results. Plasma shape parameters (minor radius and plasma elongations) share half of the net electricity power sensitivity for the DEMO 2015 design while the toroidal magnetic field and the 95 % safety factor are responsible for 23% and 17% of the electric power sensitivity, respectively. The plasma minor radius is responsible for 45% of the pulse duration sensitivity for the DEMO 2015 design, while plasma physics parameters drive ∼ 37% of the pulse duration sensitivity.

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.

The chemical vapour deposition (CVD) diamond torus window unit is a sub-component of the ITER Electron Cyclotron Heating and Current Drive (EC H&CD) system used for a diverse range of applications including plasma heating and control of plasma magneto-hydrodynamic (MHD) instabilities. It consists of an ultra-low loss polycrystalline diamond disk brazed to copper cuffs and then enclosed by a metallic structure. The diamond disk with 1.11 mm thickness already passed successfully the ITER final design review (FDR) in 2018. In view of the complete window assembly FDR, prototyping activities of the window are essential and, therefore, they shall start soon in order to check the feasibility of the proposed manufacturing and assembling sequence of the component. In this perspective, as the design of the systems surrounding the window is currently in development phase, the paper describes the finite element method (FEM) analyses of the window to carry out with the aim to prove the soundness of the design used for the prototyping and also to define requirements for the surrounding systems. Specific methodologies are adopted such as the limit analysis approach for the external loads acting on the window unit. Several combinations of forces and moments are applied to the window unit to find the maximum loads, i.e. the limits loads, which generate stresses in the unit equal to the allowable ones, according to the selected design criteria.

International audience; In the framework of the European “HORIZON 2020” research program, the EUROfusion Consortium develops a design of a fusion demonstrator (DEMO). CEA-Saclay, with the support of Wigner-CR and IPP-CR, is in charge of one of the four Breeding Blanket (BB) concepts investigated in Europe for DEMO: the Helium Cooled Lithium Lead (HCLL) BB. The BB directly surrounding the plasma is a major component ensuring tritium self-sufficiency, shielding against neutrons from D-T plasmas and heat extraction for electricity conversion.In this article, the equatorial outboard module of the HCLL reference DEMO BB (“advanced-plus” BB) concept is studied regarding thermal and mechanical behavior during normal and accidental conditions that led to modify the design and enhance the performances. A numerical approach based on the Finite Element Method (FEM) is followed. The methodology established, the assumptions done as well as the results obtained in each case are reported and critically analysed, scrutinizing some open issues on this “advanced-plus” reference concept.

The conceptual design of the demonstration fusion power reactor, known as DEMO, is ongoing and several reactor configurations have to be investigated by exploring different design parameters. For these reasons, within the European framework, systems codes like SYCOMORE (SYstem COde for MOdelling tokamak REactor) have been developed. SYCOMORE includes several specific modules, one of which is aimed to define a suitable design of the helium breeding blanket. The research activity has been devoted to improve the method to define automatically the First Wall design starting from thermal-hydraulic and thermo-mechanical considerations, using analytical design formulae and, also, taking into account the design criteria coming from Codes&Standards. Thanks to these considerations, it has been possible to derive the first dimensions of the First Wall channels from which all the other characteristics are deducted. Afterwards, it has been assessed the thermal and mechanical field using a theoretical approach. Therefore, in order to compare and validate the results, a 3D geometric model has been created and FEM analyses have been carried out finding out very satisfying results with maximum error of 6.06% for the thermal analysis while a maximum error of 20% for primary and secondary stresses.

Within the framework of EUROfusion R&D activities a research campaign has been performed at CEA-Saclay, in close collaboration with the University of Palermo, in order to investigate thermal and thermomechanical performances of the “Optimized Conservative” concept of DEMO Helium-Cooled Lithium Lead breeding blanket (HCLL). Attention has been paid to the HCLL outboard equatorial module (OEM) when subjected to the steady state nominal loading scenario. To this purpose three simplified 3D models, characterized by an increasing level of detail, have been set-up taking into account, firstly, a single radial-toroidal slice, then a basic module geometric unity composed by two adjacent slices and adding, lastly, the peripheral poloidal region. This latter 3D model has allowed the assessment of the Caps potential influence on the module thermal and thermomechanical behaviour. For each investigated 3D model, thermal and thermomechanical analyses have been performed and a stress linearization procedure has been carried out in order to verify the fulfilment of the criteria prescribed by the RCC-MRx 2015 code. The study has been performed adopting a numerical approach, based on the Finite Element Method (FEM), and adopting the Siemens NX v. 10.0 software in order to discretize the geometric domain, whereas thermal and thermomechanical calculations have been carried out using the Cast3 M 2015 FEM code. The obtained results, herewith reported and critically discussed, allow predicting a good thermal and mechanical behaviour of the “Optimized Conservative” concept of DEMO HCLL OEM, even if some small modifications to the module cooling scheme should be performed in order to avoid the insurgence of hotspots where temperature is slightly above the Eurofer limit temperature (550 °C). This will entail, from the mechanical point of view, a reduction of the secondary stress amount which is the main responsible of the failure in RCC-MRx criteria verification within First Wall-Side Wall bend region.

In the frame of the European R&D effort to develop a fusion demonstrator (DEMO) a new concept of Helium Cooled Lithium Lead breeding blanket, called “Advanced-Plus” design, is studied to improve the tritium breeding. The aim of this paper is to present the nuclear analysis of this new HCLL design carried out with the TRIPOLI-4® Monte Carlo code and the Joint European Fission and Fusion nuclear data version 3.2 (JEFF-3.2). Nuclear quantities such as: neutron flux, tritium breeding ratio, nuclear heating, displacement damage and helium production are reported. Sensitivity analysis is also carried out to investigate the impact, in tritium breeding, of the next DEMO baseline with reduced outboard thickness and the advantage, in tritium production point of view, of the single module segment against the current multi-module segment option.

In Europe (EU), in the frame of the EUROfusion consortium activities, four Breeding Blanket (BB) concepts are being developed with the aim of fulfilling the performances required by a near-term fusion power demonstration plant (DEMO) in terms of tritium self-sufficiency and electricity production. The four blanket options cover a wide range of technological possibilities, as water and helium are considered as possible coolants and solid ceramic breeder in combination with beryllium and PbLi as tritium breeder and neutron multipliers. The strategy for the BB selection and operation has to account for the challenging schedule of the EU DEMO, the ambitious operational requirements of the BBs and the still large development needed to have a BB qualified and licensed for operating in DEMO. In parallel to the continuous design efforts on the four blanket concepts, their integration in-vessel and ex-vessel has started. On the one hand it has become clear that despite the numerous systems to be integrated in-vessel the protection of the blanket first wall has to be addressed with highest priority. On the other hand the ex-vessel interfaces and the requirements imposed by the blanket to the primary heat transfer system and to the PbLi loop components have a considerable impact on the whole DEMO Plant layout. The aim of this paper is: to present the strategy for the DEMO BB down selection and BB operation in DEMO; to describe the status of the design evolution of the four EU BB concepts; to provide an overview of the challenges of the in-vessel and ex-vessel integration of the main systems interfacing the BBs and describe their design status.

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.

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.

The paper overviews activities focused on qualification of EUROFER97 structural material, introduced under a probationary phase in the nuclear components design and construction code RCC-MRx, and identification/analyses of gaps in the respective material database to be filled in. Additionally the available design rules in the code are reviewed to verify their applicability to the specificities of EUROFER97 steel and to the TBM design and fabrication. Progress achieved in development of fabrication technologies and procedures applied for manufacturing of the TBM sub-components, like, HCLL and HCPB cooling plates, stiffening plates, first wall and side caps, and for TBM structure sub-assembly is described. The used technologies are based on fusion (laser and TIG) and diffusion (HIP) welding techniques taking into account specificities of the EUROFER97 steel. With help of the agreed notified body, an appropriate approach/methodology for qualification of the developed, TBMs-related preliminary welding procedure specifications has been identified and future steps established.

This paper presents the nuclear analysis of the European DEMO baseline 2015 with HCLL blanket carried out with the TRIPOLI-4® Monte Carlo code and the JEFF-3.2 nuclear data library. The TRIPOLI-4® model was imported from CAD using the McCad tool. A procedure that generates the detailed 3D model describing all the HCLL blanket internal structures was developed. This procedure allows parametrization of the blanket internal structures such as the number of cooling plates, manifolds, etc. and the thickness of the stiffening grid for instance. Different design variants were studied to improve the tritium production. From this previous study a complete nuclear analysis was carried out on a promising design which is a compromise between tritium production and mechanical robustness. All criteria (TBR, nuclear heating in coils and displacement damage in vacuum vessel) are met using this new reference design.

The helium cooled lithium lead (HCLL) blanket concept was originally developed in CEA at the beginning of 2000: it is one of the two European blanket concepts to be tested in ITER in the form of a test blanket module (TBM) and one of the four blanket concepts currently being considered for the DEMOnstration reactor that will follow ITER. The TBM is a highly optimized component for the ITER environment that will provide crucial information for the development of the DEMO blanket, but its design needs to be adapted to the DEMO reactor. With respect to the TBM design, reduction of the steel content in the breeding zone (BZ) is sought in order to maximize tritium breeding reactions. Different options are being studied, with the potential of reaching tritium breeding ratio (TBR) values up to 1.21. At the same time, the design of the back supporting structure (BSS), which is a DEMO specific component that has to support the blanket modules inside the vacuum vessel (VV), is ongoing with the aim of maximizing the shielding power and minimizing pumping power. This implies a re-engineering of the modules' attachment system. Design changes however, will have an impact on the manufacturing and assembly sequences that are being developed for the HCLL-TBM. Due to the differences in joint configurations, thicknesses to be welded, heat dissipation and the various technical constraints related to the accessibility of the welding tools and implementation of non-destructive examination (NDE), the manufacturing procedure should be adapted and optimized for DEMO design. Laser welding instead of TIG could be an option to reduce distortions. The time-of-flight diffraction (TOFD) technique is being investigated for NDE. Finally, essential information expected from the HCLL-TBM program that will be needed to finalize the DEMO design is discussed.

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.

In the EU DEMO design (Romanelli, 2012; Federici et al., 2014), due to the large number of complex systems inside the tokamak vessel it is of vital importance to address the in-vessel integration at an early stage in the design process. In the EU DEMO design, after a first phase in which the different systems have been developed independently based on the defined baseline DEMO configuration, an effort has been made to define the interface requirements and to propose the strategies for the mechanical integration of the auxiliary heating and fuelling systems into the Vacuum Vessel and the Breeding Blanket. This work presents the options studied, the engineering solutions proposed, and the issues highlighted for the mechanical in-vessel integration of the DEMO fuelling lines, auxiliaries heating systems, and diagnostics. (C) 2017 The Author(s). Published by Elsevier B.V.

In the helium-cooled lead lithium (HCLL) blanket concept convective phenomena caused by nonuniform thermal conditions due to bulk neutron volumetric heating can occur. Buoyancy can become very important and modify the velocity distribution and related heat transfer performance of the blanket. A numerical study has been performed to investigate liquid metal flows driven by buoyant forces in a breeder unit (BU) of a HCLL test blanket module (TBM) under the influence of intense uniform magnetic fields. According to the last design review, two internal cooling plates subdivide the fluid domain into three slender flow regions, which are thermally and electrically coupled through common walls. First, a uniform volumetric heat source is considered to identify the basic convective patterns that establish in the liquid metal. Results are then compared with those obtained by applying a realistic radial distribution of the power density as obtained from a neutronic analysis. This paper summarizes the main effects of spatial gradients of a neutron thermal load on velocity and temperature distribution in magnetohydrodynamic flows in a BU of a HCLL TBM.

International audience; The design of a DEMO reactor requires the design of a blanket system suitable of reliable T production and heat extraction for electricity production. In the frame of the EUROfusion Consortium activities the Breeding Blanket Project has been constituted in 2014 with the goal to develop concepts of Breeding Blankets for the EU PPPT DEMO; this includes integrated design and a RandD program with the goal to select after 2020 concepts on fusion plants for the engineering phase. The design activities are presently focalized around a pool of solid and liquid breeder blanket with helium, water and PbLi cooling. Development of tritium extraction and control technology, as well manufacturing and development of solid and PbLi breeders are part of the Programme.

International audience; The EUROfusion Consortium develops a conceptual design of a fusion power demonstrator (DEMO) in the framework of the European Horizon 2020 innovation and research program. Key issues for DEMO are tritium self-sufficiency and heat removal for conversion into electricity. These functions are filled by the breeding blankets surrounding the plasma chamber. The Helium Cooled Lithium Lead (HCLL) blanket is one of the concepts which are investigated for DEMO. It uses the liquid lithium lead eutectic as tritium breeder and neutron multiplier and helium gas as coolant for both the Eurofer structure and the breeder. Within the EUROfusion organisation CEA is in charge for the design of the HCLL blanket for DEMO including the nuclear design analyses.This paper presents results of the nuclear analysis carried out in 2014 at CEA with the TRIPOLI-4 Monte Carlo code. In a previous analysis of the HCLL DEMO, conducted by ENEA within the EFDA work programme 2013, the MCNP code was applied with a detailed 3D Monte Carlo model describing the HCLL blanket modules (stiffening grid, cooling plate, manifold, back plates, etc.). This MCNP model was converted into a TRIPOLI-4 geometry model for performing the nuclear analysis at CEA with the objective to demonstrate consistency between the MCNP and TRIPOLI results. A very good agreement was obtained for all of the relevant nuclear responses (neutron wall loading, tritium breeding ratio, nuclear heating, etc.), showing the validity of CEA's nuclear analysis approach for the European DEMO based on TRIPOLI-4.

The Helium Cooled Lithium Lead Test Blanket Module (HCLL-TBM) is one of the two European TBMs, representative of DEMO blanket technology, that shall be ready for installation in ITER starting from the first day of operation. It is developed by the CEA in the framework of the activities of the Test Blanket Module Consortium of Associates and under contract with the European Joint Undertaking Fusion For Energy (F4E). A series of 4 different HCLL-TBMs will be installed into one of the ITER equatorial ports during the experimental campaign. This paper aims at presenting the development process and design status of HCLL-TBM, with particular focus on the so-called “integral-TBM (IN-TBM)”, which will include all the main features of the corresponding DEMO blanket module. The design of such a complex component already poses considerable issues and is complicated by the peculiarities of the ITER environment (high temperature, radiation damage, pulsed plasma operation), which are not always addressed in present nuclear design codes. After recalling the main features of the HCLL IN-TBM, supporting design analyses are presented and manufacturing techniques are discussed. © 2011 Elsevier B.V. All rights reserved.

Among the test blanket modules (TBMs) to be tested in the successive ITER phases, European design activities have focused on the electromagnetic (EM) TBM to be tested in ITER first phase (H-H), and on the integral (IN) TBM to be tested in the full duty D-T phase. The scope of the present contribution is to report activities performed on the helium cooled lithium lead (HCLL) TBM, in its EM version for the H-H phase. This phase is characterized for the TBM by relevant D-T phase magnetic field, surface heat flux and disruption induced loads, and the lack of nuclear volumetric heating. On one hand, H-H phase gives the opportunity of experiments and use of instrumentation which will not be possible in the irradiated environment of D-T phase. On the other hand, H-H phase specificity is a temperature level in the breeder zone much lower than that in the D-T phase. Main EU objectives assigned to TBM testing in H-H phase for design validation and for the build-up of a reliability database can be fulfilled at this temperature level, which is not the case for the scientific experimental program (MHD, corrosion and tritium permeation). Hence, partial heating of the TBM is foreseen. A technical solution for an electrically heated configuration of the breeder zone has been investigated. Finite element (FE) analyses were performed to assess the thermal behaviour of the TBM for the non-heated and heated cases.

Eurofer97 is a Reduced Activation Ferritic-Martensitic (RAFM) steel developed for use as structural material in fusion power reactors blankets and in particular the future DEMOnstration power plant that should follow ITER. In order to evaluate the performances of the different blanket concepts in a fusion-relevant environment, the ITER experimental programme foresees the installation of dedicated Test Blanket Modules (TBMs), representative of the corresponding DEMO blankets, in selected equatorial ports. To be fully relevant, TBMs will have to be designed and fabricated using DEMO relevant technologies and will, in particular, use Eurofer97 as structural material. While the use of ferritic/martensitic steels is not new in the nuclear industry, the fusion environment in ITER poses new challenges for the structural materials. Besides, contrary to DEMO, ITER is characterised by a strongly pulsed mode of operation that could have severe consequences on the lifetime of the components. This paper gives an overview of the issues related to the design of Eurofer97 structures in TBM components, discussing the choice of reference Codes&Standards and the consistency of the design rules with Eurofer97 mechanical properties.

This paper concerns the design calculations and performance evaluation of the helium cooled lithium lead-test blanket module (HCLL-TBM) for ITER. Previous works have shown the importance of integrated multi-physics approaches for enabling correct TBM design and specifications. Particularly, this work presents the results of a transient thermo-fluid/thermal-stress analysis in which the thermo-hydraulical and the thermo-mechanical calculations are sequentially and recursively coupled. The above procedure has been used to study the thermo-mechanical behaviour of the HCLL-TBM taking into account the transient loading conditions due to the pulsed plasma operation in ITER. New finite elements (FE) models have been specifically developed for this task: all calculations have been carried out in three dimensions with representative CAD models, including detailed descriptions of all relevant TBM components. Temperature evolutions in the coolant, structural and breeder materials have been calculated. The results of the FE calculations have been analyzed with respect to thermo-hydraulics and thermo-mechanical stresses in order to verify the compliance with the established design criteria (maximum temperature in the structural material and RCC-MR rules with SDC-IC complements).

Test Blanket Modules (TBM) operation shall not impact the ITER machine safety/availability. The monitoring of all the features relevant for the control and safe operation of the systems will therefore be indispensable. Under this perspective, the control and monitoring parameters to be measured in the Helium Cooled Lithium Lead (HCLL) TBM system are summarized and tabulated in this paper. Relevant sensors’ range and operating conditions are indicated. Their environmental and geometrical constraints in the TBM and its systems are reported. The most suitable sensors and/or measuring techniques are then suggested. Required characteristics are checked against off-the-shelf ones, when available, and needed development for their adaptation to the TBM environment is identified.

During manufacturing, nominal fabrication dimensions of the Helium Cooled Lithium Lead (HCLL). Test Blanket Module (TBM) will be machined within certain ranges and these deviations could affect the thermo-hydraulic and mechanical design criteria. In this paper, a generic methodology is presented to assess the impact on the maximum allowable temperature in the FW of 3 geometrical design factors, the distance from the plasma facing side of the first wall (FW) to the cooling channel and the poloidal and radial dimensions of the cooling channel. This methodology is sufficiently generic to be extended to other factors or to other design criteria. It mainly consists in building a surrogate model following saturated fractions of a 3n factorial design (Rechtschaffer factorial design). In that purpose, numerical experiments assuming 10% of deviation with respect to the nominal dimensions have been run using a 3D finite elements thermal model. The exploitation of the surrogate model allows quasi straight fully to assess: (a) Acceptable margins for a manufacturing technique (prior to the fabrication of the part). (b) The global effect of the final defects on the manufactured part. This surrogate model is to be considered as a straight and useful tool to select a manufacturing process or to reject a manufacture.

The Helium-Cooled Lithium-Lead and the Helium-Cooled Pebble Bed are the two breeding blankets concepts for the DEMO reactor which have been selected by EU to be tested in ITER in the framework of the Test Blanket Module projects. They both use a 9%CrWVTa Reduced Activation Ferritic–Martensitic steel, called EUROFER, as structural material and helium as coolant. This paper gives an overview of the status of the EUROFER qualification program and discusses the future needs for design criteria requirements and fabrication validation.

This paper concerns the latest design of the Helium Cooled Lithium-Lead (HCLL)-Test Blanket Module (TBM) according to the new dimensions of the space available for the module in the ITER equatorial ports. The performances of the proposed design are evaluated by means of new FE models of one breeding unit and the corresponding box structure. Detailed temperature and stress field distributions in each part of the structure have been obtained for nominal and accidental load conditions. The results of the FE calculations are analyzed with respect to thermal-hydraulics and thermo-mechanical stresses in order to verify the compliance with the established design criteria (maximum temperature in the structural material and SDC-IC rules). Also, the operating parameters are discussed in view of the DEMO-relevance of the proposed TBM design. Finally, possible improvements and modifications of the current design are proposed.

This paper gives an overview of the most recent developments for the Helium-Cooled Lithium Lead Test Blanket Modules (HCLL-TBM) in terms of TBM design, related analyses, fabrication developments and safety features. It also addresses the issues concerning the interfaces of the HCLL-TBM system with ITER and the corresponding proposals of its integration in the ITER machine and buildings. Beside the overview of the progresses realized in several domains of this project, the paper finally outlines the remaining R&D necessary for the main unsolved issues to cope with an installation of the HCLL-TBM system for day one of ITER operation.

The present design of the helium-cooled lithium-lead (HCLL) blanket foresees the successive cooling of three sub-components (the first wall, the stiffening grid, and the breeder cooling plates) by parallel He-cooling channels. The optimisation of the coolant flow is complicated by the fact that the heat recovered by the cooling elements depends on the coolant parameters (temperature, velocity, heat exchange coefficient). Since all these parameters are intrinsically related, the classical approach of obtaining the coolant temperature distribution by enthalpy balance and using this distribution as an input for finite element analyses cannot be used. In this paper we present new finite elements models of the blanket that allow to obtain the temperature distribution in the coolant on the basis of the detailed power density distribution in the surrounding structures. At the same time, the models allow to improve the description of the heat transfer between the solid walls and the coolant. A review of the most used correlations available in literature to estimate the heat transfer coefficient has then been performed and the most appropriate one has been selected. The new models have been used to carry out a thermal–hydraulic analysis of the cooling system of the HCLL breeding blanket. Moreover, an estimation of the pressure drops in the different cooling elements has been made. The obtained results are discussed in terms of differences with those obtained in previous analyses and of the consequences on the blanket design.

Silicon carbide composites are the primary composite materials being evaluated and developed world-wide for fusion structural applications. Their use as structural material involves however the adoption of appropriate design methodologies capable to take into account their complex non linear mechanical behavior. Some behavioral models developed specifically for SiCf/SiC composites and already available in literature are reviewed in this paper. They have been evaluated with respect to the possibility of implementation in FEM codes and keeping in mind fusion-specific issues. One of them has recently been implemented in the FEM code CASTEM 2000. At the same time, appropriate resistance criteria are necessary to correctly assess the potential advantages of SiCf/SiC composites over more common metallic materials. A new resistance criterion has then been defined and is presented in this work. Examples of application of the improved design analysis methodology are given throughout the paper.

TAURO is a self-cooled Pb17Li blanket for Fusion Power Reactors (FPRs) using SiCf/SiC composites structures. It has been developed with the objective to achieve both passive safety and high thermal efficiency. The Pb17Li outlet temperature exceeds 850 °C. A Brayton cycle is envisaged for power conversion and the possibility of hydrogen production is addressed. This paper recalls the main features of the TAURO blanket, and presents the performed investigations on a range of potential operations conditions with regard to plant thermal efficiency and H2-production. It also presents the recent advances made in the design methodologies of composite structures for fusion components involving the implementation in finite-element codes of material models capable of describing the complex non-linear mechanical behavior of these composites and the definition of new resistance criteria. It is possible with this new model to better treat the thermo-mechanical results. Indeed, stresses in each direction are compared directly with allowable limits for tension, compression and shear stresses in each orthotropic direction. Such a model has been implemented in the FEM code CASTEM 2000 developed at CEA. Fabrication aspects are discussed based on a brazing process currently under development and briefly presented. Moreover, the advantages of using newly-produced advanced fibers in SiCf/SiC composites in relation with the main issues (namely irradiation stability and thermal conductivity) are assessed. The potential of a liquid metal cooled divertor using SiCf/SiC as structural material, in line with the TAURO reactor characteristics is also investigated. A circuit in series could be envisaged, where the Pb17Li flows at first in the divertor and then it passes in the blanket, which allows increasing the blanket inlet temperature. Calculations on a castellated square tube basic concept showed a reasonable maximum allowable heat flux of about 5 MW/m2.

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.

Tin–lithium alloys have several attractive thermo-physical properties, in particular high thermal conductivity and heat capacity, that make them potentially interesting candidates for use in liquid metal blankets. This paper presents an evaluation of the advantages and drawbacks caused by the substitution of the currently employed alloy lead–lithium (Pb–17Li) by a suitable tin–lithium alloy: (i) for the European water-cooled Pb–17Li (WCLL) blanket concept with reduced activation ferritic–martensitic steel as the structural material; (ii) for the European self-cooled TAURO blanket with SiCf/SiC as the structural material. It was found that in none of these blankets Sn–Li alloys would lead to significant advantages, in particular due to the low tritium breeding capability. Only in forced convection cooled divertors with W-alloy structure, Sn–Li alloys would be slightly more favorable. It is concluded that Sn–Li alloys are only advantageous in free surface cooled reactor internals, as this would make maximum use of the principal advantage of Sn–Li, i.e., the low vapor pressure.

Within the framework of EUROfusion R&D activities, CEA-Saclay has carried out an investigation of the thermal and mechanical performances of alternative designs intended to enhance the Tritium Breeding Ratio (TBR) of the Helium-Cooled Lithium Lead (HCLL) Breeding Blanket (BB) for DEMO. Neutronic calculations performed on the 2014 DEMO HCLL baseline predicted a value of TBR equal to 1.07, lower than the required value of 1.1, necessary to ensure the tritium self-sufficiency of the breeding blanket taking into account uncertainties. In order to reach the TBR target, the strategy of the steel amount reduction inside the HCLL module breeding zone (BZ) has been followed by suppressing some stiffening/cooling plates inside the BZ, leading to this “advanced” concept. Since all the plates inside the BZ are actively cooled by helium, each change in their geometric layout has a strong impact on the thermal response of the module. Moreover, the removal of stiffening plate may impact the resistance of the box in case of in-module loss of coolant accident (LOCA). A thermal and mechanical campaign of analyses has been carried out in order to assess a potentially optimized layout of the module which could comply with the whole set of rules foreseen for the HCLL BB design. To perform this research campaign, a theoretical-numerical approach, based on the Finite Element Method (FEM), has been followed and the qualified Cast3M and NX FEM codes have been adopted. Results obtained are herewith reported and critically discussed, highlighting the open issues and suggesting the pertinent modifications to DEMO HCLL module design.

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.

A large part of the gyrotron R&D activities at Karlsruhe Institute of Technology focus on addressing challenging requirements posed on gyrotrons by the European concept for a demonstration fusion power plant (EU DEMO). This paper reports on the progress of these activities and on the recent results. Extensive theoretical and experimental investigations are ongoing at Karlsruhe Institute of Technology (KIT) within the Work Package Heating and Current Drive (WPHCD), coordinated by the Power Plant Physics and Technology (PPPT) Department of EUROfusion. The focus is on gyrotron R&D, which is necessary for meeting the needs of the envisaged Electron Cyclotron (EC) heating and current drive system for a European demonstration fusion power plant (EU DEMO) [1].

The European Community is committed to the development of a DEMOnstration fusion power plant whose operation could start as soon as 2050. The blanket is one of the most critical components in a fusion reactor; three of the four blanket concepts currently under development are based on the use of the liquid eutectic alloy Pb-15.7Li. Since the blanket will operate under the strong magnetic field used to confine the plasma, electromagnetic forces will occur in the PbLi flow, giving rise to magnetohydrodynamic (MHD) phenomena