Thermal Response of Bonded CFC/OFHC Divertor Mock-Ups for Fusion Experimental Reactors under Large Numbers of Cyclic High Heat Loads

Plasma facing components for fusion experimental reactors such as the ITER/FER will be exposed to severe heat loads under high heat flux and large number of thermal cycles. From the engineering point of view, divertor mock-ups with different armor tile materials have been prepared in order to investigate their overall performances, in particular an adhesive property between the armor tile and the heat sink metal. Thermal cycling tests of the divertor mock-ups have been carried out under ITER/FER relevant heat flux conditions in a particle beam engineering facility at JAERI. As results of the tests, it has been confirmed that boned carbon-fiber-composite/copper (CFC/OFHC) divertor mock-ups have resisted to 10.0 MW/m² for 1,000 cycles without cracks. At this experimental condition, the integrity and the durability of the bonds have also been confirmed. Furthermore, some bonded CFC/OFHC samples have resisted to 12.5 MW/m² for 1,000 cycles without increase of the surface temperature, although a small crack was observed at a corner of the bonded layer. Residual stress from brazing has also been analyzed for three-dimensional models. The analytical results were not different in the results of manufactured test samples that no cracks or detachments in many samples were observed.     

Design and evaluation of an optimized W/Cu interlayer for W monoblock components

Divertor plasma-facing components of future fusion reactors should be able to withstand heat fluxes of 10-20 MW/m² in stationary operation. Tungsten blocks with an inner cooling tube made of CuCr1Zr, so-called monoblocks, are potential candidates for such water-cooled components. To increase the strength and reliability of the interface between the W and the cooling tube of a Cu-based alloy (CuCr1Zr), a novel advanced W-fibre/Cu metal matrix composite (MMC) was developed for operation temperatures up to 550 °C. Based on optimization results to enhance the adhesion between fibre and matrix, W fibres (W_f) were chemically etched, coated by physical vapour deposition with a continuously graded W/Cu_PVD interlayer and then heated to 800 °C. The W_f/Cu MMC was implemented by hot-isostatic pressing and brazing process in monoblock mock-ups reinforcing the interface between the plasma-facing material and the cooling channel. The suitability of the MMC as an efficient heat sink interface for water-cooled divertor components was tested in the high heat flux (HHF) facility GLADIS. Predictions from finite element simulations of the thermal behaviour of the component under loading conditions were confirmed by the HHF tests. The W_f/Cu MMC interlayer of the mock-ups survived cyclic heat loads above 10 MW/m² without any damage. One W block of each tested mock-up showed stable thermal behaviour at heat fluxes of up to 10.5 MW/m².     

Effect of stationary high heat flux and transient ELMs-like heat loads on the divertor PFCs

The experimental evaluation of the divertor plasma facing components (PFCs) lifetime under transient events, such as edge localized modes (ELMs) and high heat flux (HHF) thermal fatigue expected during ITER normal operations and slow transient events is here presented. The experiments have been performed in the frame of an EU/RF collaboration. For carbon fiber composite material the erosion is caused by PAN fiber damage whilst the erosion of tungsten is determined by the melt layer movement and crack formation. The conclusion of this study is that, in addition to the structural change produced in the armor materials by ELMs-like loads, some mock ups showed also a degradation of the thermal fatigue performances     

Behaviour of Ti-doped CFCs under thermal fatigue tests

In spite of the remarkable progress in the design of in-vessel components for the divertor of the first International Thermonuclear Experimental Reactor (ITER), a great effort is still put into the development of manufacturing technologies for carbon armour with improved properties. Newly developed 3D titanium-doped carbon fibre reinforced composites and their corresponding undoped counterparts were brazed to a CuCrZr heat sink to produce actively cooled flat tile mock-ups. By exposing the mock-ups to thermal fatigue tests in an electron beam test facility, the material behaviour and the brazing between the individual constituents in the mock-up was qualified. The mock-ups with titanium-doped CFCs exhibited a significantly improved thermal fatigue resistance compared with those undoped materials. The comparison of these mock-ups with those produced using pristine NB31, one of the reference materials as plasma facing material for ITER, showed almost identical results, indicating the high potential of Ti-doped CFCs due to their improved thermal shock resistance.     

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.     

Numerical Analysis of Fatigue Behavior of ITER-Like Monoblock Divertor Interlayer Under Coupled Heat Loads

High-confinement mode is a very prominent operation style for future fusion device due to its unique advantages. However, the conjuncted edge localized modes (ELMs) are very difficult to control so that divertor plates are very prone to suffer both stationary high heat flux (HHF) loads of long-pulse operating mode and transient shock loads of ELMs. Most previous researches focus on degradation of plasma facing material (PFM), however, as a layer joining PFM and cooling tube, the soft copper interlayer suffers concentrated thermal stress loads due to mismatched thermal expansion of PFM and cooling tube. Its thermal fatigue behavior under such coupled loads is also of great significance to structural safety of divertor component. With such a motivation, the reduction effects on fatigue life time of a typical interlayer of monoblock divertor under series of coupled HHF and ELMs shock loading conditions are investigated. It is found that: (1) The transient shock feature of ELMs loading is propagated into interlayer with less sharp pattern. The increase of damage induced by coupled ELMs loading is limited in single cycle, while the accumulated damage of multiple consecutive coupled loading cycles is increased nonlinearly. (2) Under the coupled HHF and ELMs loading, the fatigue life time of interlayer is generally decreasing. The magnitude of decrease is increasing nonlinearly with the magnitude of ELMs peak and averaged heat flux. (3) For three characteristic parameters of ELMs shock loading such as frequency, duration and peak heat flux, the peak heat flux and frequency are two parameters more sensitive to determine coupled reduction effects on fatigue lifetime of the interlayer, while for high frequency case, time averaged heat flux takes the lead.     

Numerical Study of High Heat Flux Performances of Flat-Tile Divertor Mock-ups with Hypervapotron Cooling Concept

The hypervapotron (HV), as an enhanced heat transfer technique, will be used for ITER divertor components in the dome region as well as the enhanced heat flux first wall panels. W-Cu brazing technology has been developed at SWIP (Southwestern Institute of Physics), and one W/CuCrZr/316LN component of 450 mm×52 mm×166 mm with HV cooling channels will be fabricated for high heat flux (HHF) tests. Before that a relevant analysis was carried out to optimize the structure of divertor component elements. ANSYS-CFX was used in CFD analysis and ABAQUS was adopted for thermal–mechanical calculations. Commercial code FE-SAFE was adopted to compute the fatigue life of the component. The tile size, thickness of tungsten tiles and the slit width among tungsten tiles were optimized and its HHF performances under International Thermonuclear Experimental Reactor (ITER) loading conditions were simulated. One brand new tokamak HL-2M with advanced divertor con?guration is under construction in SWIP, where ITER-like ?at-tile divertor components are adopted. This optimized design is expected to supply valuable data for HL-2M tokamak.     

Impact of plastic softening of over-aged CuCrZr alloy heat sink tube on the structural reliability of a plasma-facing component

Precipitation-hardened CuCrZr alloy is used in fusion experiments as heat sink material for water-cooled plasma-facing components. When exposed to long-term high-heat-flux (HHF) plasma operation, CuCrZr will undergo over-ageing and thus plastic softening. In this situation, the softened CuCrZr heat sink tube will suffer from substantial plastic straining and thus fatigue damage in the course of the cyclic HHF loads. In this paper, a computational case study is presented regarding the cyclic plasticity behaviour of the over-aged CuCrZr cooling tube in a water-cooled tungsten mono-block divertor component. Finite element analysis was performed assuming ten typical HHF load cycles and using the Frederick-Armstrong constitutive equation together with corresponding material parameters. It was shown that plastic shakedown and low cycle fatigue (LCF) would be caused in the heat sink tube when softening of CuCrZr should occur. On the other hand, neither elastic shakedown nor cumulative plastic strain (ratchetting) was found. LCF design life of the CuCrZr tube was estimated based on the ITER materials handbook considering both hardened and softened states of CuCrZr. Substantial impact of softening of the CuCrZr alloy on the LCF lifetime of the heat sink tube was demonstrated.     

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.     

Thermo-mechanical analysis of the Wendelstein 7-X divertor

The stellarator Wendelstein 7-X (W7-X) has a divertor consisting of 10 units installed inside the plasma vessel (PV). It was decided not to install the long pulse high-heat flux (HHF) divertor targets at the first two years stage of W7-X operation and to start with an adiabatically cooled test divertor unit (TDU) and shorter plasma pulses operation. This allows to accumulate operation experience with much simpler components, and as a result to adjust accurately the actively cooled HHF divertor which replaces the TDU for the stationary operation. Finite element (FE) analyses have been performed for better understanding of thermo-mechanical problems of divertor targets, and to guide the design of the TDU and HHF divertors. This paper presents the detailed results of the temperature response, the deformation and thermal stress of the divertor components.     

Mechanical analysis of the joint between Wendelstein 7-X target elements and the divertor frame structure

The target elements of the actively cooled high heat flux (HHF) divertor of Wendelstein 7-X are made of CFC (carbon fiber-reinforced carbon composite) tiles bonded to a CuCrZr heat sink and are mounted onto a support frame. During operation, the power loading will result in the thermal expansion of the target elements. Their attachment to the support frame needs to provide, on the one hand, enough flexibility to allow some movement to release the induced thermal stresses and, on the other hand, to provide enough stiffness to avoid a misalignment of one target element relative to the others. This flexibility is realized by a spring element made of a stack of disc springs together with a sliding support at one of the two or three mounting points. Detailed finite element calculations have shown that the deformation of the heat sink leads to some non-axial deformation of the spring elements. A mechanical test was performed to validate the attachment design under cyclic loading and to measure the deformations typical of the expected deformation of the elements. The outcome of this study is the validation of the design selected for the attachment of the target elements, which survived experimentally the applied mechanical cycling which simulates the thermal cycling under operation.     

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 flux testing of mock-ups for a full tungsten ITER divertor

In order to evaluate the option to start the ITER operation with a full tungsten (W) divertor, the EU-DA launched an extensive R&D program. It consisted in its initial phase in the high heat flux (HHF) testing of W mock-ups and medium scale prototypes up to 20 MW/m² in the AREVA FE 200 facility (F). Critical heat flux (CHF) experiments were carried out on the items which survived the above thermal fatigue testing.After 1000 cycles at 10 MW/m², the full W Plasma Facing Components (PFCs) mock-ups successfully sustained either 1000 cycles at 15 MW/m² or 500 cycles at 20 MW/m².However, some significant surface melting, as well as the complete melting of a few monoblocks, occurred during the HHF thermal fatigue testing program representative of the present ITER requirements for the strike point region, namely 1000 cycles at 10 MW/m² followed by 1000 cycles at 20 MW/m².The results of the CHF experiments were also rather encouraging, since the tested items sustained heat fluxes in the range of 30 MW/m² in steady-state conditions.     

Flow Field and Thermal Analysis of the Divertor Target Plate for HL-2A Tokamak

In the initial phase of the physics experiment, the double-null divertor plates used consist of graphite armor tiles, Mo-alloy intermediate layers and Cu-alloy coolant tubes. In the later operating phase, tungsten will be used as armor tiles. A multi-physical field numerical analysis method is used in this paper. Its analysis model reflects more realistically the real divertor structure than other models. Two-dimensional （2D） and three-dimensional （3D） fluid flow field, temperature distribution and thermal stress analyses of the divertor plates are carried out by the ANSYS code. During the physics experimental phase with a heat flux of 1 MW/m2, a coolant velocity of 5.48 m/s, and a thermal stress of 750 kg/cm2, the graphite armor tiles successfully meet the requirements of temperature, thermal stress and sputtering erosion. The tungsten armor will be considered as a second candidate. The result of simulation can be used for upgrading the design parameters of the HL-2A poloidal divertor.     

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.     

Qualification and post-mortem characterization of tungsten mock-ups exposed to cyclic high heat flux loading

In order to evaluate the option to start the ITER operation with a full tungsten (W) divertor, high heat flux tests were performed in the electron beam facility FE200, Le Creusot, France. Thereby, in total eight small-scale and three medium-scale monoblock mock-ups produced with different manufacturing technologies and different tungsten grades were exposed to cyclic steady state heat loads. The applied power density ranges from 10 to 20 MW/m² with a maximum of 1000 cycles at each particular loading step. Finally, on a reduced number of tiles, critical heat flux tests in the range of 30 MW/m² were performed.Besides macroscopic and microscopic images of the loaded surface areas, detailed metallographic analyses were performed in order to characterize the occurring damages, i.e., crack formation, recrystallization, and melting. Thereby, the different joining technologies, i.e., hot radial pressing (HRP) vs. hot isostatic pressing (HIP) of tungsten to the Cu-based cooling tube, were qualified showing a higher stability and reproducibility of the HIP technology also as repair technology. Finally, the material response at the loaded top surface was found to be depending on the material grade, microstructural orientation, and recrystallization state of the material. These damages might be triggered by the application of thermal shock loads during electron beam surface scanning and not by the steady state heat load only. However, the superposition of thermal fatigue loads and thermal shocks as also expected during ELMs in ITER gives a first impression of the possible severe material degradation at the surface during operational scenarios at the divertor strike point.     

Examination of C/C flat tile mock-ups with hypervapotron cooling after high heat flux testing

Two C/C flat tile mock-ups with a hypervapotron cooling concept, have been successfully tested beyond ITER specification (3000 cycles at 15 MW/m², 300 cycles at 20 MW/m² and 800-1000 cycles at 25 MW/m²) in two electron beam testing facilities [F. Escourbiac, et al., Experimental simulation of cascade failure effect on tungsten and CFC flat tile armoured HHF components, Fusion Eng. Des., submitted for publication; F. Escourbiac, et al., A mature industrial solution for ITER divertor plasma facing components: hypervapotron cooling concept adapted to Tore Supra flat tile technology, Fusion Eng. Des. 75-79 (2005) 387-390]. Both mock-ups provide a SNECMA SEPCARB^® NS31 armour, which has been joined onto the CuCrZr heat sink by active metal casting (AMC) and electron beam welding (EBW). No tile detachment or sudden loss of single tiles has been observed; a cascade-like failure of flat tile armours was impossible to generate. At the maximum cyclic heat flux load of 25 MW/m² all tested tiles performed well except one, which revealed already a clear indication in the thermographic examination at the end of the manufacture. Visual examination and analysis of metallographic cuts of the remaining tiles demonstrated that the interface has not been altered. In addition, the shear strength of the C/C to copper joints measured after the high heat flux (HHF) test has been found to be still above the interlamellar shear strength of the used C/C material. The high resistance of the interface is explained by a modification of the C/C to copper joint interface due to silicon originating from the used C/C material.     

High heat load properties of actively cooled W/CuCrZr mock-ups by diffusion bonding with Ni or Ti interlayer

Two actively cooled mock-ups with 5 mm thick tungsten armor, joined to CuCrZr alloy, were successfully developed by diffusion bonding technique with Ti or Ni interlayer for the EAST device in ASIPP. Its thermal response and thermal fatigue properties were investigated with active cooling. No cracks and voids occurred at the interface of W/CuCrZr after thermal response test with a heat flux from 0 MW/m² to 10 MW/m². It survived up to 200 cycles under 10 MW/m². The temperature distributions of the mock-up were estimated by Finite Element Analysis. The simulation results indicated that thermal contact capability between the tungsten and the copper alloy with Ti interlayer was higher than that of Ni interlayer. Results showed that diffusion bonding of W/CuCrZr with Ni or Ti interlayer is a potential candidate for a high heat resistance armor material on plasma facing components (PFC).     

Development of the armoring technique for ITER Divertor Dome

This paper describes the current status of the technique for armoring of Plasma Facing Units (PFUs) of the ITER Divertor Dome with flat tungsten tiles planned for application at the procurement stage.Application of high-temperature vacuum brazing for armoring of High Heat Flux (HHF) plasma facing components was traditionally developed at the Efremov Institute and successfully tried out at the ITER R&D stage by manufacturing and HHF testing of a number of W- and Be-armored mock-ups [1], [2]. Nevertheless, the so-called “fast brazing” technique successfully applied in the past was abandoned at the stage of manufacturing of the Dome Qualification Prototypes (Dome QPs), as it failed to retain the mechanical properties of CuCrZr heat sink of the substrate. Another problem was a substantially increased number of armoring tiles brazed onto one substrate. Severe ITER requirements for the joints quality have forced us to refuse from production of W/Cu joints by brazing in favor of casting.These modifications have allowed us to produce ITER Divertor Dome QPs with high-quality tungsten armor, which then passed successfully the HHF testing. Further preparation to the procurement stage is in progress.     

Current status of He-cooled divertor development for DEMO

A He-cooled divertor concept for DEMO is being investigated at the Forschungszentrum Karlsruhe within the framework of the EU power plant conceptual study. The design goal is to resist a heat flux of 10 MW/m² at least. The major R&D areas are design, analyses, fabrication technology, and experimental design verification. A modular design is preferred for thermal stress reduction. The HEMJ (He-cooled modular divertor with multiple-jet cooling) was chosen as reference concept. It employs small tiles made of tungsten, which are brazed to a thimble made of tungsten alloy W-1%La₂O₃. The W finger units are connected to the main structure of ODS Eurofer steel by means of a copper casting with mechanical interlock. The divertor modules are cooled by helium jets (10 MPa, 600 °C) impinging onto the heated inner surface of the thimble.In cooperation with the Efremov Institute a combined helium loop & electron beam facility (60 kW, 27 keV) was built in St. Petersburg, Russia, for experimental verification of the design. It enables mock-up testing at a nominal helium inlet temperature of 600 °C, an internal pressure of 10 MPa, and a pressure difference in the mock-up of up to 0.5 MPa. Technological studies were performed on manufacturing of the W finger mock-ups. Several high heat flux tests were successfully performed till now. Post-examination and characterisation of the mock-ups subjected to the high heat flux tests were performed in collaboration with Forschungszentrum Jülich. Altogether, the test results confirm the divertor performance required. The helium-cooled divertor concept was demonstrated to be feasible. The knowledge gained from these experiments and some aspects on the design improvement are discussed in this contribution.