Computational thermo-fluid exploratory design analysis for complex ITER first wall/shield components (original) (raw)
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The Design of Actively Cooled Plasma-Facing Components
Physica Scripta, 2001
In future fusion devices, like in the stellarator Wendelstein 7-X, the target plates of the divertor will be exposed to heat loads up to power densities of 10 MW/m 2 for 1000 s. For this purpose actively cooled target elements with an internal coolant £ow return, made of 2-D CFC armor tiles brazed onto a two tube cooling structure were developed and manufactured at the Forschungszentrum JÏlich. Individual bent-and coolant £ow reversal elements were used to achieve a high £exibility in the shape of the target elements. A special brazing technology, using a thin layer of plasma-arc deposited titanium was used for the bonding of the cooling structure to the plasma facing armor (PFA). FEM-simulations of the thermal and mechanical behavior show that a detachment of about 25% of the bonded area between the copper tubes and the PFA can be tolerated, without exceeding the critical heat £ux at 15 MW/m 2 or a surface temperature of 1400 C at 10 MW/m 2 by using twisted tape inserts with a twist ratio of 2 at a cooling water velocity of 10 m/s. Thermal cycling tests in an electron beam facility up to a power density level 10.5 MW/m 2 show a very good behavior of parts of the target elements, which con¢rms the performance under fusion relevant conditions. Even defected parts in the bonding interface of the target elements, known from ultrasonic inspections before, show no change in the thermal performance under cycling, which con¢rms also the structural integrity of partly defected regions.
ITER plasma facing components, design and development
Fusion Engineering and Design, 1991
The paper summarizes the collaborative effort of the ITER Conceptual Design Activity (CDA) on Plasma Facing Components (PFC) which focused on the following main tasks: (a) the definition of basic design concepts for the First Wall (FW) and Divertor Plates (DP), (b) the analysis of the performance and likely lifetime of these PFC designs including the identification of major critical issues, (c) the start of R&D work giving already first results, and the definition of the required further R&D program to support the contemplated ITER Engineering Design Activity (EDA).
Design of the ITER EDA plasma facing components
Fusion Engineering and Design, 1998
The design of the plasma facing components (PFC) in ITER has evolved with the detailed design of the reactor. The structures exposed to the plasma have different requirements according to their functions. The primary wall, surrounding most of the plasma along the last closed magnetic surface, is exposed to a moderate heat flux (0.5 MW m − 2 ) but has to withstand the highest neutron load. The baffle wall is exposed to a peak heat flux of 3 MW m − 2 and to severe erosion from neutral particles due to their high neutrals pressure in the divertor. The limiter is subjected to the same loads as the primary wall during plasma burn conditions and a higher peak heat flux (depending on its location) during the start-up and shut down phases when the plasma is leaning on its surface. The divertor vertical targets intercept the open magnetic flux surfaces near the separatrix and have to withstand the highest heat flux and erosion in their lower part. The divertor dome is located directly below the null point and works in conditions similar to the baffle. The divertor wings receive similar thermal loads as the dome but can be subjected to high heat shocks and electromagnetic forces during plasma disruption. The paper describes the solutions adopted for the PFC and the results of analyses performed to validate the design. The description is focused on the part of the PFC which is exposed to the plasma.
Plasma facing and high heat flux materials – needs for ITER and beyond
Journal of Nuclear Materials, 2002
Plasma-facing materials (PFMs) have to withstand particle and heat loads from the plasma and neutron loads during reactor operation. For ITER knowledge has been accumulated by operation experience and dedicated tests in present tokamaks as well as by laboratory experiments and modelling. The rationale for the selection of PFMs in ITER (Be, W, carbon fibre reinforced carbon) is described with regard to the critical issues concerning PFMs, esp. erosion during transient heat loads and the T-inventory in connection with the redeposition of carbon. In the fusion reactor generation after ITER the very stringent conditions of increased surface power to be removed from the plasma, a lifetime requirement of several operational years, high neutron fluences and increased operation temperature are exerting even more severe constraints on the selection of possible materials. Comparing these boundary conditions with materials under development and their further potential, only a narrow path is left regarding heat sink and PFMs. In this context the investigations on W as first wall material carried out e.g. in ASDEX Upgrade are being discussed as well as laboratory results on W-based material systems. The implications of these results are the starting point of what should form a consistent programme towards plasma-facing and heat sink materials for a fusion reactor.
Fusion Engineering and Design, 2016
The European DEMO power reactor is currently under conceptual design within the EUROfusion Consortium. One of the most critical activities is the engineering of the plasma-facing components (PFCs) covering the plasma chamber wall, which must operate reliably in an extreme environment of neutron irradiation and surface heat and particle flux, while also allowing sufficient neutron transmission to the tritium breeding blankets. A systems approach using advanced numerical analysis is vital to realising viable solutions for these first wall and divertor PFCs. Here, we present the system requirements and describe bespoke thermo-mechanical and thermo-hydraulic assessment procedures which have been used as tools for design. The current first wall and divertor designs are overviewed along with supporting analyses. The PFC solutions employed will necessarily vary around the wall, depending on local conditions, and must be designed in an integrated manner by analysis and physical testing.
2021
The ITER tokamak, the experimental fusion reactor designed to be the first to produce net energy, has had a monoblock concept selected for use as a plasma facing component in the divertor region. This design currently consists of a CuCrZr cooling pipe surrounded by a copper interlayer and embedded in a tungsten armour plate. Additive manufacturing may facilitate a geometry capable of greater efficiency through the introduction of greater design freedom whilst maintaining compatibility with the monoblock concept. This is achieved through the addition of high conductivity material to the armour domain surrounding the coolant pipe. Finite element simulation of the heat transfer system combined with a topology optimisation methodology has been used to find the optimal distribution of high thermal conductivity material (such as Cu) for three thermal objectives: minimising temperature and thermal gradient, and maximising conductive heat flux. The topology optimisation relies on a density-...
European development of prototypes for ITER high heat flux components
Fusion Engineering and Design, 2000
The extensive EU research and development, on international thermonuclear experimental reactor (ITER) high heat flux (HHF) components aims at the demonstration of prototypes for the divertor and baffle with challenging operating requirements. The recent progress of this development is summarised in the paper, particularly concerning the manufacture and testing of mock-ups and prototypes. The available results demonstrate the feasibility of robust solutions with carbon and tungsten armour.
Status of fabrication development for plasma facing components in the EU
Fusion Engineering and Design, 2002
This paper summarises the European R&D efforts for the manufacture of shield modules and divertor cassettes for the International Thermonuclear Experimental Reactor (ITER), including their plasma facing components. The various development steps are described as they had to be taken to resolve the fabrication issues, and to keep track with the evolving design requirements and solutions. For all components, the manufacturing feasibility has been demonstrated on prototype scale which puts Europe in the position to start the procurement as soon as the decision about ITER construction is taken. The time period remaining until then is used to optimise the fabrication processes and to develop more cost effective alternatives.
Progress of detailed design and supporting analysis of ITER thermal shield
2010
The detailed design of ITER thermal shield (TS), which is planned to be procured completely by Korea, has been implemented since 2007. In this paper, the design and the supporting analysis are described for the critical components of the TS, the vacuum vessel TS (VVTS) outboard panel, labyrinths and VVTS supports. The wall type of VVTS outboard panel was changed from double wall to single wall, and the verification analyses were carried out for this design change. The dimensions of the labyrinths were determined and the heat load through the labyrinth was analyzed to check the design requirement. The preliminary result of the VVTS inboard and outboard supports were obtained considering the structural rigidity.
Overview of decade-long development of plasma-facing components at ASIPP
Nuclear Fusion, 2017
The first EAST (Experimental Advanced Superconducting Tokamak) plasma ignited in 2006 with non-actively cooled steel plates as the plasma-facing materials and components (PFMCs) which were then upgraded into full graphite tiles bolted onto water-cooled copper heat sinks in 2008. The first wall was changed further into molybdenum alloy in 2012, while keeping the graphite for both the upper and lower divertors. With the rapid increase in heating and current driving power in EAST, the W/Cu divertor project was launched around the end of 2012, aiming at achieving actively cooled full W/Cu-PFCs for the upper divertor, with heat removal capability up to 10 MW m −2. The W/Cu upper divertor was finished in the spring of 2014, consisting of 80 cassette bodies toroidally assembled. Commissioning of the EAST upper W/ Cu divertor in 2014 was unsatisfactory and then several practical measures were implemented to improve the design, welding quality and reliability, which helped us achieve successful commissioning in the 2015 Spring Campaign. In collaboration with the IO and CEA teams, we have demonstrated our technological capability to remove heat loads of 5000 cycles at 10 MW m −2 and 1000 cycles at 20 MW m −2 for the small scale monoblock mockups, and surprisingly over 300 cycles at 20 MW m −2 for the flat-tile ones. The experience and lessons we learned from batch production and commissioning are undoubtedly valuable for ITER (International Thermonuclear Experimental Reactor) engineering validation and tungsten-related plasma physics.