One current design of the divertor for a fusion reactor like DEMO uses He-cooled thimble-like tungsten, which is covered by sacrificial tungsten tile. Each thimble has to be connected with a ...supporting unit made from ferritic steel. This paper describes the development of joining techniques between tungsten thimbles and steel supporting units. Paper also provides an evaluation of simple geometries up to more complex conical interlocks filled with cast copper. Four candidates tungsten alloys (WL10, W-single crystal, W–Cu composite and chemical vapour deposited (CVD) tungsten) were experimentally checked by ‘non-isothermal’ heating to characterize the thermal gradient in the range 600
°C (for joint) and more than 1000
°C (for thimble top) using a special testing procedure. Basing on the test results, several mock-ups were manufactured for future high heat flux testing in a helium loop.
At the Forschungszentrum Karlsruhe (FZK), designs of He-cooled divertor concepts are pursued for near-term reactor models like DEMO. Due to the new concept presented here, small structured tungsten ...parts are no longer needed for the design. In combination with new assembly and sub-module designs this results in promising characteristics regarding cooling performance, reliability and feasibility. The cooling method and conceptual design of the He-jet-cooled divertor shall be presented in this study, together with an assessment of the performance based on computational fluid dynamics (CFD) simulations.
Within the EU power plant conceptual study (PPCS), a modular He-cooled divertor concept is being investigated at the Forschungszentrum Karlsruhe to achieve a heat flux of at least 10
MW/m
2. The ...intermediate-term goal of divertor development is the completion of a test divertor module (TDM) which is envisaged to be tested in ITER from 2020 onward before it will be used in DEMO. As a preparatory step, the possibility of performing such an experiment with a helium-cooled TDM in ITER is assessed. The investigation covers, e.g. checking whether the space available is sufficient and the experiment is compatible with ITER operation and RH (remote handling) procedures as well as assessing the thermohydraulic and piping layout for helium cooling.
Within the EU power plant conceptual study (PPCS), a modular He-cooled divertor concept (Ref. 1) has been investigated at the Forschungszentrum Karlsruhe to achieve a heat flux of at least 10 MW/m
2
.... The divertor conceptual design is based on the use of a tile made of tungsten, a structural element made of tungsten alloy, and a steel cartridge. The cooling of the divertor module is realized by an impingement of helium jets (10 MPa, 600 °C) flowing through an array of small jet holes located at the top of the cartridge, able to remove the high heat flux incident on the top surface of the tiles.
In this paper a modular design of a helium cooled divertor is introduced. A method of design examination regarding the cooling capability and the component stresses are pointed out. The method is based on the use of a combined system of modern computer tools. For the 3D design construction, the CAD program CATIA V5 was utilized. The simulation calculations were performed in two steps: thermo-hydraulic CFD calculations using the ANSYS CFX tool and thermo-mechanical FEM calculations with the ANSYS code. The CFD computations were done taking into account the design geometry with an appropriate meshing and the boundary conditions, i.e. the defined heat flux, the helium pressure and temperature at the inlet. Among other things, the heat-transfer-coefficients received from the CFD runs were then used for the following FEM analyses. The simulation results and a potential of design improvement will be discussed.
Gas cooled divertor concepts are regarded as a suitable option for fusion power plants because of an increased thermal efficiency for power conversion systems and the use of a coolant compatible with ...all blanket systems. A modular helium cooled divertor concept is proposed with an improved heat transfer. The concept employs small tiles made of tungsten and brazed to a finger-like structure made of Mo-alloy (TZM). Design goal was a heat flux of at least 15 MW/m
2 and a minimum temperature of the structure of 600
°C. The divertor has to survive a number of cycles (100–1000) between operating temperature and room temperature even for the steady state operation assumed. Thermo-hydraulic design requirements for the concepts include to keep the pumping power below 10% of the thermal power to the divertor plates, and simultaneously achieving a heat transfer coefficient in excess of 60 kW/m
2 K. Inelastic stress analysis indicates that design allowable stress limits on primary and secondary (thermal) stresses as required by the ITER structural design criteria are met even under conservative assumptions. Finally, critical issues for future development are addressed.
Strong impulses to the development of in-vessel components for near-term fusion reactors like DEMO were given by the recent EU power plant conceptual study (PPCS). Within the PPCS reactor models were ...developed based on the EU Helium cooled pebble bed blanket (HCPB) concept (reactor model B) and the dual coolant blanket (DC) concept (reactor model C). As a consequence of the study, a design review was carried out in the EU to create a modular HCPB blanket, followed by an effort at Forschungszentrum Karlsruhe (FZK) of a complete in-vessel integration. Also, the development of a gas cooled divertor was launched within the PPCS, with the aim of increasing both safety and the overall plant efficiency. The latest and most advanced divertor concept, which was developed at FZK, is based on a feasible and effective heat transfer enhancement technique, namely the multiple jet impingement cooling technology, while helium was chosen as a gaseous coolant due to its good safety characteristics when used with a Be-containing blanket. A brief description of the divertor design is given with the focus is on reactor integration and target layout.
A helium-cooled divertor concept for future power plants to be build after ITER is investigated at the Forschungszentrum Karlsruhe
1. Simulations by computational fluid dynamics (CFD) programs show ...that the divertor is capable of removing a heat load of 10
MW/m
2 at least at an acceptable pumping power.
To validate the simulation results, experiments in the helium loop, Helium Blanket Test Loop (HEBLO), are run under conditions less severe than the operational ones in the fusion reactor. The results of these test campaigns are presented and compared to the simulations. The agreement is good.
First wall and cooling plates are considered the most important structural parts of the EU HCPB blanket concept which is based on the use of ferritic–martensitic steel as structural material, Li
4SiO
...4 pebbles as breeder material, beryllium pebbles as neutron multiplier, and 8 MPa helium as coolant. Both the first wall and cooling plates contain complex arrays of internal He coolant channels. The favourite manufacturing technology is diffusion welding of two halves of plates applying the hot isostatic pressure (HIP) welding method that allows uniform distribution of the pressure acting on the outer surfaces of the welding objects. The HIP experiment was started with small MANET specimens with internal coolant channels. The objective of this work is to investigate the appropriate HIP technique, boundary conditions, and parameters in order to achieve good mechanical properties of the welding joints as well as to achieve a transition to test specimens of larger dimensions.
The Forschungszentrum Karlsruhe develops for more than 20 years fusion technologies and structural materials for ITER and ultimately for DEMO. While the technologies developed for ITER have reached ...some maturity, although not yet completed, the technologies and structural materials for DEMO need substantial progress before such a commercial prototype fusion reactor can be built. The main advances required beyond the technologies already developed for ITER are: the development of high temperature superconducting coils, a He or liquid metal cooled breeding blanket, a He-cooled divertor and the integration of the “In Vessel” components into a DEMO (Tokamak) reactor (a complex design task). In addition the development of suitable structural materials for the blanket and the divertor is an urgent task. These structural materials have to be of a low activation kind (decay to low activation level within ∼100 years) and have to achieve a suitable lifetime (∼5 years, i.e. ∼150
dpa) when used for DEMO blankets and divertor and when considering the large neutron affluence in DEMO. In all these fields the Forschungszentrum Karlsruhe is very active and in many cases a world leader.
All the recent DEMO design studies for helium cooled divertors utilize tungsten materials and alloys, mainly due to their high temperature strength, good thermal conductivity, low erosion, and ...comparably low activation under neutron irradiation. The long-term objective of the EFDA fusion materials programme is to develop structural as well as armor materials in combination with the necessary production and fabrication technologies for future divertor concepts. The programmatic roadmap is structured into four engineering research lines which comprise fabrication process development, structural material development, armor material optimization, and irradiation performance testing, which are complemented by a fundamental research programme on “Materials Science and Modeling”. This paper presents the current research status of the EFDA experimental and testing investigations, and gives a detailed overview of the latest results on fabrication, joining, high heat flux testing, plasticity, modeling, and validation experiments.
