Progress of He-cooled divertor development for DEMO

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/m² 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/m² unscathed.     

Experimental and numerical investigation of a full-scale helium-cooled divertor finger mock-up

A modular helium-cooled divertor design based on the multi-jet impingement concept (HEMJ) that is capable of accommodating a surface heat flux of at least 10 MW/m² has been developed at the Forschungszentrum Karlsruhe. Experimental investigations with a full-scale mock-up designed and built at the Georgia Institute of Technology, Atlanta were carried out in the helium loop HEBLO at Karlsruhe. Tests were run at heat loads of up to 2 MW/m² and flow conditions of 38 °C and 8 MPa so that the Reynolds number matches that for the actual divertor operating conditions (21,600). Comparison between the experimental data and results of simulations performed using the computational fluid dynamics code ANSYS CFX showed good agreement for the cooled surface temperature distribution, while the pressure loss was underestimated by about 20% by the code.     

Optimizing the overall configuration of a He-cooled W-alloy divertor for a power plant

A number of different He-cooled divertor configurations have been proposed for magnetic fusion energy (MFE) power plant application. They range in scale from a plate configuration with characteristic dimension of the order of 1 m, to the ARIES-CS T-tube configuration with characteristic dimension of the order of 10 cm, to the EU FZK finger concept with characteristic dimension of the order of 1.5 cm. All these designs utilize tungsten or tungsten alloy as structural material. This paper considers the characteristics of the different divertor configurations and proposes the possibility of optimizing the design by combining different configurations in an integrated design based on the anticipated divertor heat flux profile.     

Non-destructive examination of the bonding interface in DEMO divertor fingers

Plasma facing components (PFCs) with tungsten (W) armor materials for DEMO divertor require a high heat flux removal capability (at least 10 MW/m² in steady-state conditions). The reference divertor PFC concept is a finger with a tungsten tile as a protection and sacrificial layer brazed to a thimble made of tungsten alloy W – 1% La₂O₃ (WL10). Defects may be located at the W thimble to W tile interface. As the number of fingers is considerable (>250,000), it is then a major issue to develop a reliable control procedure in order to control with a non-destructive examination the fabrication processes. The feasibility for detecting defect with infrared thermography SATIR test bed is presented. SATIR is based on the heat transient method and is used as an inspection tool in order to assess component heat transfer capability. SATIR tests were performed on fingers integrating or not the complex He cooling system (steel cartridge with jet holes). Millimeter size artificial defects were manufactured and their detectability was evaluated. Results of this study demonstrate that the SATIR method can be considered as a relevant non-destructive technique examination for the defect detection of DEMO divertor fingers.     

Powder Injection Molding – An innovative manufacturing method for He-cooled DEMO divertor components

At Karlsruhe Institute of Technology (KIT), a He-cooled divertor design for future fusion power plants has been developed. This concept is based on the use of modular cooling fingers made from tungsten and tungsten alloy, which are presently considered the most promising divertor materials to withstand the specific heat load of 10 MW/m². Since a large number of the finger modules (n > 250,000) are needed for the whole reactor, developing a mass-oriented manufacturing method is indispensable. In this regard, an innovative manufacturing technology, Powder Injection Molding (PIM), has been adapted to W processing at KIT since a couple of years. This production method is deemed promising in view of large-scale production of tungsten parts with high near-net-shape precision, hence, offering an advantage of cost-saving process compared to conventional machining.The complete technological PIM process for tungsten materials and its application on manufacturing of real divertor components, including the design of a new PIM tool is outlined and, results of the examination of the finished product after heat-treatment are discussed. A binary tungsten powder feedstock with a solid load of 50 vol.% was developed and successfully tested in molding experiments. After design, simulation and manufacturing of a new PIM tool, real divertor parts are produced. After heat-treatment (pre-sintering and HIP) the successful finished samples showed a sintered density of approximately 99%, a hardness of 457 HV0.1, a grain size of approximately 5 μm and a microstructure without cracks and porosity.     

High heat flux test with HIP bonded Be/Cu/SS mock-ups for the ITER first wall

In order to verify the integrity of the first wall (FW) of the International Thermonuclear Experimental Reactor (ITER), especially for preparing its qualification program by ITER-O, Be/Cu/SS mock-ups, which were the same size as the qualification mock-ups, were fabricated and tested at the TSEFEY, an e-beam facility, in Efremov, Russia. These mock-ups were joined with a 316 L austenitic stainless steel (SS316L) block for a structural material, CuCrZr for a heat sink material and SS316L tubes for a coolant and then, joined with three Be tiles for an armor material. A hot isostatic pressing (HIP) was used as manufacturing methods at a 1050 °C, 100 MPa for 2 h for a Cu/SS joining and at a 580 °C, 100 MPa for 2 h for a Be/(Cu/SS) joining. Two mock-ups were fabricated by using 1 μmCr/10 μmCu of an interlayer between the Be tile and Cu block. The high heat flux (HHF) tests were performed at 1.5 and 2.0 MW/m² heat fluxes for each mock-up. The given conditions and the expected fatigue lifetime were evaluated from a preliminary analysis with ANSYS. Both mock-ups survived for up to 1000 and 268 cycles at 1.5 and 2.0 MW/m² heat fluxes, respectively. They are higher than the expected numbers of cycles to a failure.     

High heat load tests on W/Cu mock-ups and evaluation of their application to EAST device

Tungsten has been considered as the primary candidate plasma-facing materials (PFM) for the EAST device. Three actively cooled W/Cu mock-ups with an interlayer made of tungsten-copper alloy (1.5 mm) were designed and manufactured. The tungsten armors, pure sintered tungsten plate (1 mm) and plasma-sprayed tungsten coatings (0.3 and 0.9 mm), were bonded to the interlayer by brazing and depositing respectively. All mock-ups can withstand high heat flux up to 5 MW/m² and no obvious failure was found after tests. The thermal performance experiments and microstructure analyses indicated the structure of mock-ups possess good thermal contact and high heat transfer capability. WCu alloy as an interlayer can largely reduce the stress due to the mismatch and improve the reliability. The mock-up with 0.9 mm coating had the highest surface temperature than the other two mock-ups, delaminations of this mock-up were found in the near surface by SEM. The primary results show that pure sintered tungsten brazed to WCu alloy is a possible way, and thick plasma-sprayed coating technique still need to be improved.     

Design and performance study of the helium-cooled T-tube divertor concept

The ARIES-CS study has been launched with the goal of developing through physics and engineering optimization an attractive power plant concept based on a compact stellarator configuration. The study included an effort to characterize the divertor location and corresponding heat load distribution, and to develop a He-cooled divertor concept that could accommodate a heat flux of at least 10 MW/m², and that would integrate well with the other power core components. This paper describes the design study of this divertor concept, which, although developed for a compact stellarator, is well suited for a tokamak configuration also.     

Fabrication of Be/CuCrZr/SS mock-ups for ITER first wall

Three types of the mock-ups consisting of Be, CuCrZr and stainless steel (SS) were fabricated by a HIP joining method to demonstrate the fabricability of the ITER first wall. The effects of the mock-up design and the interlayer type on the joining strength of the Be/CuCrZr joint were investigated on the basis of the shear test results for the Be/CuCrZr joint specimens. The joining strength of the Be/CuCrZr ranged from 78 MPa to 204 MPa depending on the types of mock-ups and interlayers. The highest shear strength was obtained from the Be/CuCrZr joint specimens which were taken from the Be/CuCrZr/SS mock-ups with three 80 mm × 80 mm × 10 mm Be tiles and a Cr/Cu interlayer. The effect of the size and the number of Be tiles used in the mock-ups was not as significant as the interlayer selection. Even though the shear test is considered to be very useful to elucidate which interlayer is the most applicable for the joining of Be and CuCrZr, the final interlayer selection should be based on the results obtained from the high heat flux test for the mock-ups fabricated in this work.     

Influence of multiple jet cooling on the heat transfer and thermal stresses in DEMO divertor cooling finger

The divertor concept for DEMO fusion reactor is based on modular design cooled by multiple impinging jets. Such divertor should be able to withstand a surface heat flux of at least 10 MW/m² at an acceptable pumping power. To reduce the thermal loads the plasma-facing side of the divertor is build up of numerous small cooling fingers. Each cooling finger is cooled by an array of jets blowing through the holes on the steel cartridge.The size, number and arrangement of jets on the cartridge influences the heat transfer and pressure drop characteristics of the divertor. Five different cartridge designs are analyzed in the paper. The most critical parameters, such as structure temperature, heat removal ability, pressure drop, cooling efficiency and thermal stress loadings in the cooling finger are predicted for each cartridge design. A combined computational fluid dynamics and structural model was used to perform the necessary numerical analyses. The results have shown that the cartridge design with the best heat transfer and pressure drop characteristics is not also the most favorable choice from the point of view of minimum stress peaks.     

Fabrication and high heat flux test of the first wall mock-ups for the Korean He Cooled Test Blanket (KO HCML TBM)

Korea has proposed and designed a Helium Cooled Molten Lithium (HCML) Test Blanket Module (TBM) to be tested in the International Thermonuclear Experimental Reactor (ITER). Ferrite Martensitic (FM) steel is designed to be used as a structural material in this design. Three mock-ups, especially for the first wall channels, were fabricated with a Hot Isostatic Pressing (HIP, 1050 °C, 100 MPa, 2 h) in order to develop the fabrication technology. One of them was used to observe the microstructure and it was found that there are no pores and cracks. Other mock-ups were used with high heat flux (HHF) tests performed with 20 cycles under 0.5 and 1.0 MW/m² heat fluxes in order to evaluate the integrity of the fabricated mock-ups. Here, HHF test conditions were determined with an ANSYS-CFX analysis. And then, the mock-ups were tested and broken under a 1.5 MW/m² heat flux, which is about the Critical Heat Flux (CHF) at the wall.     

Evaluation of heat transfer by sublimation for the application to the divertor heat sink for high fusion energy conversion

Thermal and structural responses of divertor target were evaluated by using finite element method. High heat flux simulating ELMs at the level of 100 MW/m² was assumed onto the tungsten armor, and surface temperature profile was obtained. When dynamic heat load over 100 MW/m² was applied, the maximum surface temperature exceeded 1300 °C, and it caused recrystallization of tungsten regardless of the heat transfer below it. The result was used to conduct dynamic heat load experiment on tungsten, and material behavior of tungsten was evaluated under dynamic heat load. This study also proposed new concept of divertor heat sink which can distribute high heat flux and transfers the heat to high temperature medium. It consists of tungsten armor, composite enhanced with high thermal conductivity fiber, and heat transport system applying phase transition. High heat flux simulating ELMs was also applied to target surface of the divertor, temperature gradient, thermal stress of tungsten and composite were evaluated. Based on the results of analysis, thermal structural requirement was considered.     

Conceptual design and heat transfer performance of a flat-tile water-cooled divertor target

The divertor target components for the Chinese fusion engineering test reactor (CFETR) and the future experimental advanced superconducting tokamak (EAST) need to remove a heat flux of up to ～20 MW m-2.In view of such a high heat flux removal requirement,this study proposes a conceptual design for a flat-tile divertor target based on explosive welding and brazing technology.Rectangular water-cooled channels with a special thermal transfer structure (TTS)are designed in the heat sink to improve the flat-tile divertor target's heat transfer performance(HTP).The parametric design and optimization methods are applied to study the influence of the TTS variation parameters,including height (H),width (W*),thickness (T),and spacing (L),on the HTP.The research results show that the flat-tile divertor target's HTP is sensitive to the TTS parameter changes,and the sensitivity is T ＞ L ＞ W* ＞ H.The HTP first increases and then decreases with the increase of T,L,and W* and gradually increases with the increase of H.The optimal design parameters are as follows:H =5.5 mm,W* =25.8 mm,T =2.2 mm,and L =9.7 mm.The HTP of the optimized flat-tile divertor target at different flow speeds and tungsten tile thicknesses is studied using the numerical simulation method.A flat-tile divertor mock-up is developed according to the optimized parameters.In addition,high heat flux (HHF)tests are performed on an electron beam facility to further investigate the mock-up HTP.The numerical simulation calculation results show that the optimized flat-tile divertor target has great potential for handling the steady-state heat load of 20 MW m-2 under the tungsten tile thickness＜5 mm and the flow speed ≥7 m s-1.The heat transfer efficiency of the flat-tile divertor target with rectangular cooling channels improves by ～ 13％ and ～30％ compared to that of the flat-tile divertor target with circular cooling channels and the ITER-like monoblock,respectively.The HHF tests indicate that the flat-tile divertor mock-up can successfully withstand 1000 cycles of 20 MW m-2 of heat load without visible deformation,damage,and HTP degradation.The surface temperature of the flat-tile divertor mock-up at the 1000th cycle is only ～930 ℃.The fiat-tile divertor target's HTP is greatly improved by the parametric design and optimization method,and is better than the ITER-like monoblock and the fiat-tile mock-up for the WEST divertor.This conceptual design is currently being applied to the engineering design of the CFETR and EAST flat-tile divertors.     

Deep drawing of tungsten plates for structural divertor applications in future fusion devices

The reference design of a helium cooled divertor for future fusion reactors makes use of hundreds of thousands of finger units consisting of a pressurized structural part called a thimble. Due to the high number of parts needed, the thimble has to be fabricated by mass production techniques like deep drawing. As the thimble is a pressurized part exposed to an internal pressure of 100 bar, the demands for the material are high, which means that it requires the best available tungsten material. Former work has shown that pure tungsten material has the best impact properties and has to be preferred over other commercially available tungsten materials, such as that doped with potassium or strengthened with oxides like lanthanum oxide.Furthermore the inherent weakness of the grain boundaries has to be taken into account, which requires the need for grains that are aligned to the contour of the part (grain boundary alignment).This paper describes the successful deep drawing of a 1 mm tungsten plate in high vacuum at 600 °C. In doing this, a thimble can be machined with grains that follow the contour. Furthermore the characterization of a 1 mm tungsten plate is conducted by tensile tests at room temperature and at 600 °C, as well as by Charpy tests taking into account the anisotropic material behaviour.     

Microchannel cooling technique for dissipating high heat flux on W/Cu flat-type mock-up for EAST divertor

As an important component of tokamaks, the divertor is mainly responsible for extracting heat and helium ash, and the targets of the divertor need to withstand high heat flux of 10 MW m⁻² for steady-state operation. In this study, we proposed a new strategy, using microchannel cooling technology to remove high heat load on the targets of the divertor. The results demonstrated that the microchannel-based W/Cu flat-type mock-up successfully withstood the thermal fatigue test of 1000 cycles at 10 MW m⁻² with cooling water of 26 l min⁻¹, 30 °C (inlet), 0.8 MPa (inlet), 15 s power on and 15 s dwell time; the maximum temperature on the heat-loaded surface (W surface) of the mock-up was 493 °C, which is much lower than the recrystallization temperature of W (1200 °C). Moreover, no occurrence of macrocrack and 'hot spot' at the W surface, as well as no detachment of W/Cu tiles were observed during the thermal fatigue testing. These results indicate that microchannel cooling technology is an efficient method for removing the heat load of the divertor at a low flow rate. The present study offers a promising solution to replace the monoblock design for the EAST divertor     

Numerical studies on helium cooled divertor finger mock up with sectorial extended surfaces

Jet impinging technique is an advance divertor concept for the design of future fusion power plants. This technique is extensively used due to its high heat removal capability with reasonable pumping power and for safe operation. In this design, plasma-facing components are fabricated with numerous fingers cooled by helium jets to reduce the thermal stresses. The present study is focused towards finding an optimum performance of one such finger mock-up through systematic computational fluid dynamics (CFD) studies. Heat transfer characteristics of jet impingement have been numerically investigated with sectorial extended surfaces (SES). The result shows that addition of SES enhances heat removal potential with minimum pumping power. Detailed parametric studies on critical parameters that influence thermal performance of the finger mock-up have been analyzed. Thermo-mechanical analysis has been carried out through finite element based approach to know the state of stress in the assembly as a result of large temperature gradients. It is seen that the stresses are within the permissible limits for the present design. The whole numerical simulation has been carried out using general-purpose CFD software (ANSYS FLUENT, Release 14.0, User Guide, Ansys, Inc., 2011). Benchmark validation studies have been performed against high-heat flux experiments (B. Končar, P. Norajitra, K. Oblak, Appl. Therm. Eng., 30, 697–705, 2010) and a good agreement is noticed between the present simulation and the reported results.     

Recent US activities on advanced He-cooled W-alloy divertor concepts for fusion power plants

Several advanced He-cooled W-alloy divertor concepts have been considered recently for power plant applications. They range in scale from a plate configuration with characteristic dimension of the order of 1 m, to the ARIES-CS T-tube configuration with characteristic dimension of the order of 10 cm, to the EU FZK finger concept with characteristic dimension of the order of 1.5 cm. The trend in moving to smaller-scale units is aimed at minimizing the thermal stress under a given heat load; however, this is done at the expense of increasing the number of units, with a corresponding impact on the reliability of the system. The possibility of optimizing the design by combining different configurations in an integrated design, based on the anticipated divertor heat flux profile, also has been proposed. Several heat transfer enhancement schemes have been considered in these designs, including slot jet, multi-hole jet, porous media and pin arrays. This paper summarizes recent US efforts in this area, including optimization and assessment of the different concepts under power plant conditions. Analytical and experimental studies of the concepts and cooling schemes are presented. Key issues are identified and discussed to help guide future R&D, including fabrication, joining, material behavior under the fusion environment and impact of design choice on reliability.     

Thermal Fatigue Study on the Divertor Plate Materials

Thermal fatigue property of the divertor plate is one of the key issues that governs the lifetime of the divertor plate.Taking tungsten as surface material,a small-mock-up divertor plate was made by hot isostatic press welding (HIP),A thermal cycling experiment for divertor mock-up was carried out in the vacuum,where a high-heat-flux electronic gun was used as the thermal source,A cyclic heat flux of 9MW/m^2 was loaded onto the mock-up,a heating duration of 20s was selcted,the cooling water flow rate was 80ml/s.After 1000 Cycles,the surface and the W/Cu joint of the mock-up did not show any damage,The SEM was used to analyze the microstructure of the welding joint,where no cracks were found also.     

Model validation of GAMMA code with heat transfer experiment for KO TBM in ITER

By considering the requirements for a DEMO-relevant blanket concept, Korea (KO) has proposed a He cooled molten lithium (HCML) test blanket module (TBM) for testing in ITER. A performance analysis for the thermal–hydraulics and a safety analysis for the KO TBM have been carried out using a commercial CFD code, ANSYS-CFX, and a system code, GAMMA (GAs multicomponent mixture analysis), which was developed by the gas cooled reactor in Korea. To verify the codes, a preliminary study was performed by Lee using a single TBM first wall (FW) mock-up made from the same material as the KO TBM, ferritic martensitic steel, using a 6 MPa nitrogen gas loop. The test was performed at pressures of 1.1, 1.9 and 2.9 MPa, and under various ranges of flow rate from 0.0105 to 0.0407 kg/s with a constant wall temperature condition. In the present study, a thermal–hydraulic test was performed with the newly constructed helium supplying system, in which the design pressure and temperature were 9 MPa and 500 °C, respectively. In the experiment, the same mock-up was used, and the test was performed under the conditions of 3 MPa pressure, 30 °C inlet temperature and 70 m/s helium velocity, which are almost same conditions of the KO TBM FW. One side of the mock-up was heated with a constant heat flux of 0.3–0.5 MW/m² using a graphite heating system, KoHLT-2 (Korea heat load test facility-2). Because the comparison result between CFX 11 and GAMMA showed a difference tendency, the modification of heat transfer correlation included in GAMMA was performed. And the modified GAMMA showed the strong parity with CFX 11 calculation results.     

Manufacturing and testing of a HETS module for DEMO divertor

The development of a divertor concept for fusion power plants that is able to grant efficient recovery and conversion of the considerable fraction (～15%) of the total fusion thermal power incident is deemed to be an urgent task to meet in the EU Fast Track scenario. The He-cooled conceptual divertor design is one of the possible candidates. Helium cooling offers several advantages including chemical and neutronic inertness and the ability to operate at higher temperatures and lower pressures than those required for water cooling. The HETS (high-efficiency thermal shield) concept, initially developed by ENEA for water, has been adapted for use with He as coolant. This DEMO divertor concept is based on elements joined in series and protected by a hemispheric dome; it allows an increase of thermal exchange coefficient both for high speed of gas and for “jet impingement” effects of gas coming out from the internal side of hemispheric dome. It has been calculated to be capable of sustaining an incident heat flux of 10 MW/m² when operating at 10 MPa, an inlet He temperature of 600 °C, and an outlet temperature of 800 °C. The presented activity, performed in the frame of EFDA-TW5TRP-001 task, was focused on the manufacturing of a single HETS module and on its thermal–hydraulic testing. The materials used for the HETS module manufacturing were all DEMO-compatible: W for the armor material and the hemispherical-dome, DENSIMET for the exchanger body. The testing is performed by connecting the module to HEFUS3 He loop system that is a facility able to supply the He flow to the required testing conditions: 400 °C, 4–8 MPa and 20–40 g/s. The needed incident heat flux is obtained by RF inducting equipment coupled to an inductor coil installed just over the HETS module. A CFD analysis by ANSYS-CFX was performed in order to predict the thermal–mechanical behavior of the module and a final comparison with the experimental data is required to validate the CFD results. All parameters are monitored and recorded by data acquisition system.