The Helium Cooled Pebble Bed (HCPB) Breeding Blanket (BB) is one of the 4 BB concepts being investigated in the EU for their possible implementation in DEMO. During 2014 the former “beer-box” BB ...concept based on the ITER’s HCPB Test Blanket Module suffered several design changes so as to meet the different counteracting nuclear, thermohydraulic and thermomechanical requirements. These studies evidenced that the concept is too rigid to meet the tight TBR requirements imposed for the EU DEMO (i.e. TBR≥1.10). Additionally, the complex manifold system with unbalanced helium mass flow in each of the 2 parallel cooling loops made the concept thermohydraulically complex. However, parametric studies during 2015 revealed that the HCPB concept have potential for a better nuclear performance, as well as margin for a significant simplification of the cooling internals by redefining the cooling plates and the architecture of the blanket, building a symmetric flow scheme.
This paper describes the new HCPB concept based on an integrated FW with the breeding zone thermohydraulics and helium manifold systems. The former complex manifold backplates have been compacted and integrated in the cooling plates, releasing ≈300mm of radial space that can be used now to increase breeder zone, the neutron shielding, to reinforce the Back Supporting Structure (BSS) or basically to reduce the reactor size. Detailed neutronic analyses have yielded a TBR of ∼1.20 for the baseline design. Initial analyses show a correct thermohydraulic behavior. Preliminary thermomechanical analyses also indicate that the design can potentially withstand an in-box LOCA at 9 MPa at a level C according to the RCC-MRx code. Future consolidation activities are described, which shall lead to a concept meeting the BB requirements.
•The manufacturing and joining technologies developed at KIT for helium cooled divertors are reviewed and critically discussed.•Various technologies have been pursued and further developed aiming ...divertor components with very high quality and sufficient reliability.•Very promising routes have been found for which however still R&D works are necessary.•Technologies developed are also useful for other divertor and even blanket concepts, particularly those with tungsten armor.
In the helium cooled (HC) divertor, developed at KIT for a fusion power plant, tungsten has been selected as armor as well as structural material due to its crucial properties: high melting point, very low sputtering yield, good thermal conductivity, high temperature strength, low thermal expansion and low activation. Thereby the armor tungsten is attached to the structural tungsten by thermally conductive joint. Due to the brittleness of tungsten at low temperatures its use as structural material is limited to the high temperature part of the component and a structural joint to the reduced activation ferritic martensitic steel EUROFER97 is foreseen. Hence, to realize the selected hybrid material concept reliable tungsten–steel and tungsten–tungsten joints have been developed and will be reported in this paper.
In addition, the modular design of the HC divertor requires tungsten armor tiles and tungsten structural thimbles to be manufactured in high numbers with very high quality. Due to the high strength and low temperature brittleness of tungsten special manufacturing techniques need to be developed for the production of parts with no cavities inside and/or surface flaws. The main achievement in developing the respective manufacturing technologies will be presented and discussed.
To achieve the objectives mentioned above various manufacturing and joining technologies are pursued. Their later applicability depends on the level of development including their transferability to the component. Hence, specifying design and requirements for the components of interest will determine appropriate time and criteria for selecting most promising technologies. Although the considered technologies are mainly developed for the HC divertor it is worth to note that they are also useful for other divertor and even blanket concepts, particularly those with tungsten armor.
•Progresses on the EUROfusion DCLL breeder blanket are presented.•The new DCLL version is focused in complying with in-box LOCA requirements.•Thermal-hydraulics, neutronics and mechanical ...calculations are presented.•Two new design approaches for the FCI are shown, including the effect of RIC.
The Dual Coolant Lithium Lead breeding blanket is being investigated as a candidate for the European DEMO. This blanket is based on the use of PbLi as breeder and coolant (“self-cooled breeding zone”) and high-pressure helium for cooling the structures made of EUROFER. During the first part of the project, a conceptual design of the DCLL equatorial outboard module (which supports the highest neutronic and thermal loads) has been finalized, meeting the requirements of tritium self-sufficiency and shielding. It was designed to work under normal, undisturbed operational conditions. The present work shows the current DCLL design produced to comply with the in-box LOCA requirements, including the most relevant results on neutronics, thermal-hydraulic and mechanical calculations, as well as the advances on the Flow Channel Insert development.
Within the framework of the EU power plant conceptual study (PPCS), a He-cooled divertor concept has been investigated at Forschungszentrum Karlsruhe in cooperation with the Efremov Institute. The ...design goal is to remove a high heat load of at least 10
MW/m
2. The design is based on a modular construction of cooling finger unit that helps reduce thermal stresses. The divertor finger unit, which is cooled by high pressure helium, consists of a tungsten tile as thermal shield and sacrificial layer, and a thimble made of tungsten alloy. The success of this design depends strongly on the effectiveness of the cooling technology and on the availability of appropriate structural materials such as tungsten alloy and oxide-dispersion-strengthened (ODS) steel as well as the related fabrication and joining technology. Results of this investigation are discussed in the paper.
Tungsten is being considered as a potential plasma-facing material for future fusion devices, primarily due to its low erosion rate and high heat resistance. The intrinsic problem of this material, ...the brittleness even at elevated temperatures, requires the development and assessment of new tungsten materials.
In the frame of the European material development programme for future fusion power plants W containing 5wt.% Ta as well as uniaxially forged ultra-high purity W and powder injection moulded W were assessed in a neutral beam high heat flux test facility at IPP Garching.
The investigation of the morphology modification of tungsten, occurring during heat loading using H and He particles, which simulates the expected divertor operation conditions, is indispensable in order to develop reliable plasma-facing materials. The effects seen, erosion, gas retention and cavity formation, depend on both the loading conditions and the operating temperature.
This contribution presents a comparative study of the surface morphology changes of different W materials under hydrogen and helium beam loading to surface temperatures between 1500°C (1773K) and 2000°C (2273K), using actively cooled mock-ups. Loading is performed with pure H and mixed 94% H/6% He beams (ϕ=4×1021m−2s−1), resulting in 10MWm−2 thermal load. Pulse durations of 30s are applied to achieve fluences up to 3×1025m−2 under stationary temperature conditions.
In the frame of the conceptual design phase of the EU DEMO an effort is made here:
•To define the interface requirements among systems to be integrated in the VV and the BB.•To propose the ...integration strategies for the auxiliary heating, diagnostic and fuelling systems into the VV and the BB and for the BB and divertor supporting structures.•To define a schedule for the in-vessel components integration design analyses.•To identify the 3D supporting tools.
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.
A present topic of high interest in magnetic fusion is the “gap” between near-term and long-term concepts for high heat flux components (HHFC), and in particular for divertors. This paper focuses on ...this issue with the aim of characterizing the international status of current HHFC design concepts for ITER and describing the different technologies needed in the designs being developed for fusion power plants. Critical material and physics aspects are highlighted while evaluating the current readiness level of long-term concepts, identifying the design and R&D gaps, and discussing ways to bridge them.
A He-cooled divertor concept for DEMO
1 has been developed at Karlsruhe Institute of Technology (KIT) since a couple of years with the goal of reaching a heat flux of 10
MW/m
2 anticipated for DEMO. ...The reference concept HEMJ (He-cooled modular divertor with multiple-jet cooling) is based on the use of small cooling fingers – each composed of a tungsten tile brazed to a tungsten alloy thimble – as well as on impingement jet cooling with helium at 10
MPa, 600
°C. The cooling fingers are connected to the main structure of ODS Eurofer steel by brazing in combination with a mechanical interlock. This paper reports progress to date of the design accompanying R&Ds, i.e. primarily the fabrication technology and HHF experiments. For the latter a combined helium loop and electron beam facility (200
kW, 40
keV) at Efremov Institute, St. Petersburg, Russia, has been used. This facility enables mock-up testing at a nominal helium inlet temperature of 600
°C, a pressure of 10
MPa, and a maximal pressure head of 0.5
MPa. HHF test results till now confirm well the divertor design performance. In the recent test series in early 2010 the first breakthrough was achieved when a mock-up has survived over 1000 cycles at 10
MW/m
2 unscathed.
Within the framework of the EU power plant conceptual study (PPCS), a modular He-cooled divertor concept with integrated pin array (HEMP) is being developed at the Forschungszentrum Karlsruhe. The ...design goal is to achieve a high heat flux of at least about 10–15 MW/m
2, which is proposed for a near-term reactor model like DEMO. The development and optimization of the divertor concept require a close link between the main issues: design, analyses, materials and fabrication technology, and experiments with feedbacks between them to be accounted for. Design-specific requirements on materials and fabrication issues will be discussed.
Within the EU power plant conceptual study (PPCS), a modular He-cooled divertor concept is being investigated at the Forschungszentrum Karlsruhe to reach a heat flux of at least 10
MW/m
2. The HEMJ ...(helium-cooled modular divertor with multiple-jet cooling) has been chosen as the reference design. It consists of a tile made of tungsten, a thimble made of tungsten-alloy, and a steel cartridge. The cooling is realized by impingement jets of high-pressure helium (10
MPa, 600
°C) through an array of small jet holes located at the top of the cartridge. Besides the jet cooling ability, thermal stresses in the tungsten parts induced by the high heat loads are regarded as important factors that limit the divertor performance and lifetime. Therefore, optimizing the finger geometry for stress reduction is indispensable. In this contribution a combined CAD-FEM method for the optimization of the geometry of the finger components will be outlined.
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