DEMO Blanket Research Articles

Selective Laser Sintering as Manufacturing Process for the Realization of Complex Nuclear Fusion and High Heat Flux Components

The development of fabrication technologies for ITER and DEMO Blanket concepts is an activity followed by the KIT since a long time. A variety of fabrication technologies has been developed and qualified in strong collaboration with industry. Besides the standard technologies, an activity has been launched to explore the capabilities of generative fabrication procedures such as Laser Beam Melting (LBM) and Selective Laser Sintering (SLS).To manufacture demonstrator parts for Blankets by LBM /SLS, EUROFER (a Reduced Activation Ferritic Martensitic/RAFM steel applied e.g. in ITER) has been produced as powder metallurgical product. With this material, test parts have been realized. The test program started with solid parts and simple geometries used for extraction of specimen for material qualification purpose. Later, more complex parts were fabricated to investigate the feasibility of hollow and double walled structures and components with internal channel structures. Finally, blanket relevant part segments (e.g. for the Stiffening Plates) with meandering cooling channel structures and Flow Channel Insert segment demonstration parts for the EU Helium Cooled Pebble Bed and the Dual Coolant Lithium Lead Breeder Blanket concepts for DEMO have been fabricated.First preliminary qualification activities have been concluded using test procedures applied e.g. for the qualification of welding seams such as Tensile – and Charpy tests, macro- and micro structure investigation or hardness measurement. The findings have been compared to standard material properties of EUROFER in order to quantify the fabrication results. Material properties of ~ 80% and more, compared to standard rolled EUROFER with comparable heat treatment history could be demonstrated in case of Tensile- and Yield- strength, total strain after fracture as well as energy consumption in Charpy tests.Also the joining of generatively fabricated sub-components together with conventionally fabricated EUROFER parts by Electron Beam welding has been investigated in order to test the option of the fabrication of hybrid components. These hybrid components are intended to combine parts with straight channels fabricated by Electrical Discharge Machining together with generative fabricated parts with complex structures of cooling channels (e.g. nested U-shaped flow paths) which cannot be realized using standard machining technologies.This technical note reports the first promising qualification results of generatively fabricated EUROFER parts. Also the weldability of generative fabricated parts and conventionally fabricated EUROFER has been demonstrated. Preliminary qualification results of the welding are shown, and possibilities for experimental qualifications are discussed.

Functional materials for breeding blankets—status and developments

The development of tritium breeder, neutron multiplier and flow channel insert materials for the breeding blanket of the DEMO reactor is reviewed. Present emphasis is on the ITER test blanket module (TBM); lithium metatitanate (Li2TiO3) and lithium orthosilicate (Li4SiO4) pebbles have been developed by leading TBM parties. Beryllium pebbles have been selected as the neutron multiplier. Good progress has been made in their fabrication; however, verification of the design by experiments is in the planning stage. Irradiation data are also limited, but the decrease in thermal conductivity of beryllium due to irradiation followed by swelling is a concern. Tests at ITER are regarded as a major milestone. For the DEMO reactor, improvement of the breeder has been attempted to obtain a higher lithium content, and Be12Ti and other beryllide intermetallic compounds that have superior chemical stability have been studied. LiPb eutectic has been considered as a DEMO blanket in the liquid breeder option and is used as a coolant to achieve a higher outlet temperature; a SiC flow channel insert is used to prevent magnetohydrodynamic pressure drop and corrosion. A significant technical gap between ITER TBM and DEMO is recognized, and the world fusion community is working on ITER TBM and DEMO blanket development in parallel.

Recent R&D Results on Fusion Nuclear Technology for ITER and DEMO Reactor in Japan

Several key components, such as superconducting coils, remote handling equipment, heating systems, have been designed and manufactured by JADA (Japan Domestic Agency). These activities have been carried out in accordance with the agreed schedule; in collaboration with the ITER organization and other domestic agencies. As a significant technical program using ITER, to design and to manufacture the TBS (Test blanket system), some R&D and design activities have also been conducted in Japan. Under the IFERC (the International Fusion Energy Research Center) projects of BA (Broader Approach) activities, design and R&D activities on fusion DEMO reactor have been carried out. For the DEMO R&D activity, five basic R&D subjects for a DEMO blanket system have been selected, and been studies under close collaborations between EU and JA: structure materials (RAFM steels and SiC/SiC composites), functional materials (tritium breeders and neutron multipliers), and tritium technology. From 2007, the above projects produced a set of fruitful results. A series of advanced technologies for the DEMO blanket system has also been carried out by Universities in Japan. Some significant basic R&D studies have also been carried out under US-JA collaborative program.

Thermal optimization of the Helium-Cooled Lithium Lead breeding zone layout design regarding TBR enhancement

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.

Preliminary safety analysis of LOCAs in one EU DEMO HCPB blanket module

Safety analysis for the design basis accident (DBA) is essential to support DEMO blanket concept design. It is necessary to study the pressure behaviour in the blanket and the connected systems during the loss of coolant accident (LOCA) in a blanket module, as well as the temperature evolution in the coolant flow and the associated structures. For the Helium Cooled Pebble Bed (HCPB) blanket concept (version 2014) three representative accidental sequences of LOCA have been simulated using system code MELCOR 1.8.6 for fusion. The LOCA is identified to be the failure of cooling channels in the stiffening grid, in the FW or in the breeder unit. Simulation results are discussed in this paper.

Tritium Extraction from Lithium-Lead in the EU DEMO Blanket Using Permeator Against Vacuum

The current studies on the development of the EU DEMO breeding blanket include among the options the use of liquid Lithium-Lead (17Li-83Pb) as tritium breeder (and multiplier), with different coolants. As the tritium is steadily produced in the blanket during the reactor operation, suitably efficient strategies for the Tritium Extraction System (TES) from the breeder must be developed, allowing a closed fuel cycle in situ and avoiding tritium accumulation in the machine. The Permeator Against Vacuum (PAV) appears to be one of the most promising solutions to achieve this goal. In this paper, the performance of a PAV system is studied by means of different models describing the transport of tritium in the liquid PbLi and in the metallic membrane separating it from the vacuum. The comparison of the results for different membrane materials and size of the device, for a given target efficiency, allows to optimize the PAV design, also taking into account corrosion issues. The approximations and limitations of the adopted models are also addressed.

Activation and decay heat analysis of the European DEMO blanket concepts

Demonstrating tritium self-sufficiency is an important goal of the European tokamak demonstration fusion reactor. Currently four breeder blanket concepts are being considered: the Helium Cooled Pebble Bed (HCPB), Helium Cooled Lithium-Lead (HCLL), Dual Cooled Lithium-Lead (DCLL) and Water Cooled Lithium-Lead (WCLL). Differences in materials and construction of the four breeder blanket concepts lead to differing nuclear responses. As well as affecting tritium breeding this is also of particular importance in safety analyses, such as the modelling of loss of coolant accidents, as it affects the blanket's decay heat and nuclide inventory.This paper presents and discusses analysis performed for each of the 2014 designs of the blanket concepts to ascertain the decay heat and nuclide inventory for the entire reactor. It was found that the total decay heat at short decay times for the HCLL concept (17.5MW at 1s) was between 17% and 22% lower than the HCPB, WCLL and DCLL. At longer decay times (∼100 years) it was found that the DCLL and WCLL blankets had decay heats in the region of 2–3 orders of magnitude above the HCPB and HCLL blankets. The differences noted between the blanket concepts are discussed in terms of neutron spectrum and material composition.

Design of the helium cooled lithium lead breeding blanket in CEA: from TBM to DEMO

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.

Radiological impact mitigation of waste coming from the European fusion reactor DEMO with DCLL breeding blanket

In large fusion machines, as the foreseen DEMO, the high energy neutrons produced will cause the transmutation of the interacting materials which will become a source of radioactive waste. Besides the main constituents of a material that could transmute but which presence is essential, the impurities often give rise to significant additional activation compared to the base material. Thus, once identified the nuclides generating the dominant activation products for the DCLL (Dual Coolant Lithium-Lead) concept of Breeding Blanket (BB) for the European DEMO, the objective of the study has been to determine the impurity limits for such materials in order to minimize the radiological impact of the waste produced. This will enable to consider it as Low Level Waste and dispose it in the Spanish near-surface repository of El Cabril. For both the original compositions as well as the revised one with reduced amount of impurities, suggested to mitigate the waste impact, activation calculations have been performed. Hence, total beta-gamma activity, alfa activity, specific activity for different nuclides, decay heat and surface gamma dose rate have been analyzed with reference to the IAEA and SEAFP-2 standards for waste classification and to the specific regulations of El Cabril.

Nuclear analysis of the HCLL blanket for the European DEMO

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.

Methodological approach for DEMO neutronics in the European PPPT programme: Tools, data and analyses

The methodological approach employed for the neutronics in the PPPT (Power Plant Physics and Technology) programme of EUROfusion is presented. It encompasses development works on advanced computational tools and activities related to the nuclear design and performance evaluation of the DEMO power plant including safety, maintenance, and waste management issues. Development work is conducted on Monte Carlo codes, on the CAD geometry conversion for Monte Carlo simulations, and on coupled radiation transport and activation computation systems. The role of nuclear data for reliable DEMO neutronics design analyses and uncertainty assessments is also addressed. Specific examples of nuclear analyses are presented including breeder blanket and shielding analyses for the different DEMO blanket concepts as well as related activation, decay heat and shut-down dose rate analyses.

Analysis of the WCLL European demo blanket concept in terms of activation and decay heat after exposure to neutron irradiation

This comparative paper describes the activation and decay heat calculations for water-cooled lithium-lead performed part of the EURO fusion WPSAE programme and specifications in comparison to other European DEMO blanket concepts on the basis of using a three-dimensional neutronics calculation model. Results are provided for a range of decay times of interest for maintenance activities, safety and waste management assessments. The study revealed that water-cooled lithium-lead has the highest total decay heat at longer decay times in comparison to the helium-cooled design which has the lowest total decay heat. In addition, major nuclides were identified for water-cooled lithium-lead in W armour, Eurofer, and LiPb. In addition, great attention has been dedicated to the analysis of the decay heat and activity both from the different water-cooled lithium-lead blanket modules for the entire reactor and from each water-cooled lithium-lead blanket module separately. The neutron induced activation and decay heat at shutdown were calculated by the FISPACT code, using the neutron flux densities and spectra that were provided by the preceding MCNP neutron transport calculations.

Dynamic thermal-hydraulic modelling of the EU DEMO HCPB breeding blanket cooling loops

A global, system-level thermal-hydraulic model of the EU DEMO tokamak fusion reactor is currently under development and implementation in a suitable software at Politecnico di Torino, including the relevant heat transfer and fluid dynamics phenomena, which affect the performance of the different cooling circuits and components and their integration in a consistent design. The model is based on an object-oriented approach using the Modelica language, which easily allows to preserve the high modularity required at this stage of the design. The first module of the global model will simulate the blanket cooling system and will be able to investigate different coolant options and different cooling schemes, to be adapted to the different blanket systems currently under development in the Breeding Blanket (BB) project. The paper presents the Helium-Cooled Pebble Bed (HCPB) module of the EU DEMO blanket cooling loops system model. The model is used to compare different schemes for the cooling of the different components of the HCPB BB, and to suggest improvements aimed at optimizing the pumping power required by the cooling system. The model is then used to analyse a pulsed scenario, characteristic of the EU DEMO operation.

Fabrication and characterization of advanced neutron multipliers for DEMO blanket

Prototypic pebbles with Be12V composition, which do not undergo a peritectic reaction during cooling, were fabricated and characterized, because this composition is not only unnecessary for the homogenization treatment, but also able to prevent increase of specific surface area. The results of granulation experiments indicated that the prototypic pebbles of single-phase Be12V were successfully fabricated without a homogenization treatment. The results of hydrogen generation reaction experiments showed that the prototypic pebbles with Be12V composition exhibited superior oxidation properties compared to pure Be pebbles and similar to those of as-granulated Be–Ti beryllide pebbles: as-granulated Be12V pebbles exhibited good resistance to water vapor. The results of deuterium retention experiments indicated that beryllides exhibit lower deuterium-trapping efficiency than other tested materials. Because of a small desorption from beryllides, the total retention of deuterium in Be12V was evaluated to be approximately 20% of that in pure Be.

Tritium transport in HCLL and WCLL DEMO blankets

The Helium-Cooled Lithium Lead (HCLL) and Water-Cooled Lithium Lead (WCLL) Breeding Blankets are two of the four blanket designs proposed for DEMO reactor. The study of tritium transport inside the blankets is fundamental to assess their preliminary design and safety features. A mathematical model has been derived, in a new form making makes easier to determine the most critical components as far as tritium losses and tritium inventories are concerned, and to model the tritium performance of the whole system. Two cases have been studied, the former with tritium generation rate constant in time and the latter considering a typical pulsed operation for a time span of 100h. Tritium inventories and tritium losses are the main output of the model. Tritium concentrations, inventories and losses are initially calculated and compared for the two blankets, in a reference case without permeation barriers or cold traps. A parametric study to show the behavior of the two systems when certain parameters are changed, in order to minimize inventories and/or losses, has been carried out.

Comparison over the nuclear analysis of the HCLL blanket for the European DEMO

This paper presents the comparison over the nuclear analysis of the European DEMO with HCLL blanket carried out with the TRIPOLI-4® Monte Carlo code and the JEFF-3.1.1 nuclear data library and with the MCNP5 Monte Carlo code and the FENDL-2.1 nuclear data library. The MCNP5 analysis was conducted firstly by ENEA with a detailed 3D model describing all the HCLL blanket internal structures. This MCNP5 model was converted into TRIPOLI-4® representation for performing the nuclear analysis at CEA with the objective to demonstrate consistency between both analyses. A very good agreement was obtained for all of the relevant nuclear responses (neutron wall loading, tritium breeding ratio, nuclear heating, neutron flux distribution, etc.), validating CEA’s nuclear analysis approach, based on TRIPOLI-4® Monte Carlo code and JEFF-3.1.1 nuclear data library, for the European DEMO.

Evaluation of EM loads distribution on DEMO blanket segments and their effect on mechanical integrity

This work is aimed to analyze the EM internal forces distribution on the blanket system (blankets modules and segment back supporting structure) of the EU PPPT DEMO 2015 reactor configuration. In order to validate their impact on the segment structure, an EM analysis is conducted using a simplified plasma central disruption. The calculated Lorentz’s forces distributions are then used as input for structural analyses focusing on the mechanical integrity of the segment back supporting structure. In particular, the electrical and structural assumptions used in this work are based on the HCPB blanket design developed at the Karlsruhe Institute of Technology. A preliminary assessment of the primary stresses according the design code RCC-MRx indicates the ability of the segments to resist the EM forces, where the lowest margin is given by the immediate plastic instability criterion on the inboard segment with 14%.

Transient analyses on the cooling channels of the DEMO HCPB blanket concept under accidental conditions

Helium Cooled Pebble Bed (HCPB) blanket concept is one of the DEMO (Demonstration Power Plant) blanket concepts running for the final DEMO design selection. In this paper, transient analyses on the cooling channels of the FW are carried out by means of CFD simulations for the selected accidental scenarios loss-of-coolant-accident (LOCA) and loss-of-flow-accident (LOFA). ANSYS-CFX is used for the simulations. The simulation results help to understand how fast the temperature of the FW can increase and what is the time window that is available until the temperature of the structural material reaches the design limit in order to be able to define a suitable protection strategy for the system. In view of later developments of the models, the heat transfer coefficients calculated with CFD are compared with the values predicted by two widely used correlations for turbulent pipe flows.

Objectives and status of EUROfusion DEMO blanket studies

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 an integrated design and R&D programme 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.

Recent technical progress on BA Program: DEMO activities and IFMIF/EVEDA

The Broader Approach (BA) activities consists of three major projects: the International Fusion Energy Research Center (IFERC) project, the International Fusion Materials Irradiation Facility/Engineering Validation and Engineering Design Activities (IFMIF/EVEDA) project, and the Satellite Tokamak Programme (STP, JT-60SA). These projects have been carried out to obtain basic data for the design of DEMO fusion reactor from 2007. For 8-year activities, the above projects could produce a set of fruitful results for the DEMO reactor. DEMO design activity has been conducted to build a set of DEMO design bases in accordance with a series of discussion between EU and JA. In the DEMO R&D activities, five basic R&D subjects for a DEMO blanket system have been selected, and been studies under close collaborations between EU and JA: structure materials (RAFM steels and SiC/SiC composites), functional materials (tritium breeders and neutron multipliers), and tritium technology. Some additional R&D subjects recommended by peer review comments have also been studied successfully in recent years. Regarding the IFMIF/EVEDA project, some main components of the accelerator facility been designed and tested. The validation test using EVEDA Lithium Test Loop (ELTL) was also completed successfully in October 2014.