High Heat Flux Tests Research Articles

Fibre Bragg grating sensors for fusion diagnostics: Temperature monitoring of a tungsten mono-block mock-up under high heat flux

The optimum functioning of the divertor is very important for the realisation of the long-pulse H-mode discharge of a tokamak device. This necessitates the development of a component monitoring system for its operation status including temperature and strain. This study introduces a promising technique for monitoring temperature and strain sensor systems based on the fibre Bragg grating (FBG). It was used to monitor the temperature of a W/Cu mono-block mock-up during a high heat flux (HHF) test for the first time, demonstrating its suitability as a temperature monitoring system for the tungsten divertor. Temperatures up to 1000 °C were measured under HHF using five femtosecond FBG sensors inscribed in an optic fibre encapsulated in a stainless-steel capillary tube. The FBG sensors exhibit higher sensitivity to temperature variation while maintaining the same measurement accuracy relative to the five thermocouples (TCs) in symmetrical locations. An ANSYS fluent simulation based on the finite element method yielded an output consistent with the experimental results. The temperature distribution obtained through one channel of optical fibre sensors demonstrates the advantage of FBG sensors as a multipoint measurement method. These sensors are preferred to TCs because the complex installation process and utilisation of a mass of wires are eliminated. In view of the success of this work, the installation of an FBG-based temperature monitoring system as a novel diagnostic system is planned on the upgraded lower divertor of the Experimental Advanced Superconducting Tokamak device.

Progress in developing ITER and DEMO first wall technologies at SWIP

The first wall (FW) is key component for ITER and the Chinese DEMO—the China Fusion Engineering Test Reactor (CFETR). It faces burning plasma and has a high heat flux (HHF) surface load. Critical issues and key technologies for manufacturing the ITER and CFETR FW are discussed with regards to thermal fatigue performance, Be/CuCrZr and W/reduced activation ferritic/martensitic (RAFM) steel bonding, material properties and failure mechanisms. Design improvement of the hypervapotron cooling channel and the joint interface structure was done, which increases the thermal fatigue lifetime of the enhanced heat flux (EHF) ITER FW by more than one order of magnitude as indicated by thermo-mechanical analysis. Small mock-ups and full-size EHF FW fingers were manufactured by qualified technologies in sequence. An HHF test of the small mock-ups showed that Be tile size and defects at the Be/CuCrZr interface have a great effect on the fatigue lifetime. Manufacturing tests showed a thick oxygen-free Cu interlayer could provide a good solution for the interface cracking issue. An ITER EHF FW semi-prototype was manufactured with additional two full-size finger pairs that successfully survived the demanding HHF test at 4.7 and 5.9 MW m−2. Various manufacturing technologies for joining W/RAFM steel for the CFETR FW have been studied, including a TiN coating at the interface as a tritium permeation barrier. Hot iso-static pressed W/RAFM steel joints showed a higher bonding strength than brazing joints but varied a lot with the interlayer metals. Further studies are required to optimize the technologies.

A newly designed actively water-cooled Langmuir probe for tokamak devices

Conventional Langmuir probes used in tokamak devices get damaged easily and even melt when exposed to divertor plasma with high thermal and particle loads. Recently, the Institute of Plasma Physics, Chinese Academy of Science, aimed to overcome this issue by designing and manufacturing an uncompensated tungsten/oxygen-free copper (OFC)/CuCrZr-alloy structure-based actively water-cooled Langmuir probe, which showed excellent heat removal capability in finite element analyses (FEA). The heat resistance and antiheat fatigue property of the proposed probe were verified by conducting electron beam high heat flux (HHF) tests. Results show that the maximum temperatures of the probe’s tungsten surface were approximately 445 °C and 875 °C under heat loads of 10 and 20 MW/m2, respectively. The probe successfully passed 600 cycles of 10 and 20 MW/m2 thermal fatigue HHF tests. Moreover, no obvious damages were found on the tungsten surface and W/OFC/CuCrZr joints. FEA and HHF test results proved the reliability of the newly designed uncompensated probe. In view of this progress, it is promising to apply this actively water-cooled probe in tokamak devices for plasma diagnostics.

X-ray imaging diagnostic in the high heat flux test facilities

Ensuring an accurate visualization of incident electron beam in high heat flux test facilities is a key to improve the monitoring of high-heat flux (HHF) tests carried out in view to verify the thermo-mechanical performance of components. Recent progress in X-ray imaging sensors R&D made them well-adapted for such monitoring, in particular in the case of tested components with complex shapes. This paper describes the innovative X-ray imaging diagnostic implemented in both the TSEFEY-M and IDTF facilities (Efremov Institute, Saint-Petersburg, Russia) to improve the accuracy of ITER in-vessel components HHF tests monitoring.

Conceptual design of a high heat flux testing facility in support of CFETR

A high heat flux test facility for testing the prototype of plasma facing components (PFCs) of the China Fusion Engineering Test Reactor (CFETR) will be designed and constructed to evaluate the thermal performance of the PFCs. The conceptual design of the CFETR PFCs test facility (CPTF) has been completed. It will be installed with two electron beam (EB) guns. The beam power of two EB guns is 800 kW and 60 kW respectively. The CFETR PFCs or other in-vessel components to be tested are mounted on a manipulator in vacuum vessel (VV). The manipulator is fixed on the vacuum vessel and drives the movement and rotation of the tested components. The adjustment range of the movement and rotation are ±0.5 m and ±90°. The dimension of the vacuum vessel is ∅3×4 m. Its vacuum is 10−3 Pa. All diagnostic systems for testing experiments are mounted on the vacuum vessel, such as thermometer, infrared camera and strain measurement system. Preliminary analysis has been performed to verify the structural design rationality. The simulation results indicate that current design can meet the mechanical requirements.

Optimization of W/Cu monoblock mock-up with FGM interlayer for CFETR devertor targets

Thanks to its excellent heat removal capability, good performance under high heat flux (HHF) test and higher critical heat flux (CHF) value, the water cooled monoblock type plasma facing component (PFC) will be the optimal choice for divertor targets of CFETR and future fusion reactor. The conventional ITER-like W/Cu monoblock PFC employs a soft Cu interlayer to relax thermal stress that has deteriorative irradiation effects of swelling and embrittlement by transmutation and would undergo plastic fatigue under HHF loads. To solve these issues, W/Cu monoblock with functional graded material (FGM) interlayer is designed for CFETR and future fusion reactor. The two parameters that are the number and thickness of the FGM layers are optimized via thermal and stress simulation by finite element method. The simulation result shows that FGM interlayer plays an important role in alleviating stress. And it is deduced that the optimized number and thickness of the FGM layers are 3 layers and 0.3 mm of each layer, respectively.

A 6 MW/m2 High Heat Flux Testing Facility of Irradiated Materials Using Infrared Plasma-Arc Lamps

Assessing the effect of neutron irradiation of plasma-facing materials has been challenging due to both the technical and radiological challenges involved. In an effort to address the radiological challenges, a facility was developed to conduct high heat flux testing (HHFT) of inherently small samples of neutron-irradiated materials. A new line-focus reflector was designed and fabricated at Oak Ridge National Laboratory for a plasma-arc lamp (PAL) to attain a source heat flux of 12 MW/m2. The new reflector was fabricated with two ports for monitoring specimen condition during HHFT. At the same operational conditions for PAL, the absorbed heat flux in tungsten was increased from 1.39 MW/m2 with the uniform irradiance reflector to 5.12 MW/m2 for the line-focus reflector. This fourfold increase in the heat flux, at the same PAL electrode lifetimes, enabled cost-effective facility operation for a high number of cyclic high heat flux tests. Specifically, the test section is confined to a hemispherical dome, and specimens are bolted directly to a water-cooled copper alloy rod. Temperature measurement in the PAL facility was a main challenge due to a limited line of sight. For the first time in a PAL facility operating at high heat fluxes, the specimen surface temperature was directly measured during HHFT with a pyrometer. The HHFT data, which were obtained in this upgraded PAL facility, demonstrated the facility readiness for irradiated materials.

Manufacturing, high heat flux testing and post mortem analyses of a W-PIM mock-up

Abstract In the framework of the European material development programme for fusion power plants beyond the international thermonuclear experimental reactor (ITER), tungsten (W) is an attractive candidate as plasma facing material for future fusion reactors. The selection of tungsten is owing to its physical properties such as the high melting point of 3420 °C, the high strength and thermal conductivity, the low thermal expansion and low erosion rate. Disadvantages are the low ductility and fracture toughness at room temperature, low oxidation resistance, and the manufacturing by mechanical machining such as milling and turning, because it is extremely cost and time intensive. Powder Injection Molding (PIM) as near-net-shape technology allows the mass production of complex parts, the direct joining of different materials and the development and manufacturing of composite and prototype materials presenting an interesting alternative process route to conventional manufacturing technologies. With its high precision, the PIM process offers the advantage of reduced costs compared to conventional machining. Isotropic materials, good thermal shock resistance, and high shape complexity are typical properties of PIM tungsten. This contribution describes the fabrication of tungsten monoblocks, in particular for applications in divertor components, via PIM. The assembly to a component (mock-up) was done by Hot Radial Pressing (HRP). Furthermore, this component was characterized by High Heat Flux (HHF) tests at GLADIS and at JUDITH 2, and achieved 1300 cycles @ 20 MW/m². Post mortem analyses were performed quantifying and qualifying the occurring damage by metallographic and microscopical means. The crystallographic texture was analysed by EBSD measurements. No change in microstructure during testing was observed.

Development and Performance of Tungsten-Coated Graphitic Foam for Plasma-Facing Components

High-density graphitic foam is an ideal low-Z plasma-facing material for deuterium-deuterium plasma experiments where tritium codeposition is not an issue. However, like all carbon, graphitic foam suffers from a precipitous drop in thermal conductivity at high temperatures, >600°C. To mitigate these problems, functionally graded layers of tungsten can be deposited to a thickness of 2 to 4 mm onto the plasma side of the foam using chemical vapor deposition. The graphitic foam then acts as a high-conductivity heat sink at temperatures below 600°C for the thin high-Z armor coating. The overall component weighs 18 times less than a comparable volume of tungsten and lacks the coefficient of thermal expansion joining issues between the CuCrZr tubing and the tungsten. This paper discusses the coating development and characterization and presents the results of recent plasma exposures in W7-X. It also reports on computational fluid dynamics heat transfer modeling and preparations for high heat flux testing of mock-ups. This hybrid plasma-facing component (PFC) consisting of innovative engineered materials may be a cost-effective, actively cooled solution for the divertors and other PFCs in long-pulse machines like W7-X and WEST.

Characterising the impact of castellations on the efficiency of induction heating during testing in the HIVE facility

High heat flux testing is a vital part of engineering component validation for fusion technology. The Heat by Induction to Verify Extremes (HIVE) facility is designed to provide a practical avenue for this aspect of component testing. It provides fast turnaround for smaller concepts and a cost-effective approach by utilising induction heating within a small vacuum vessel. Due to the potential complexity of induced current paths in an induction heating system, the extent and homogeneity of power coupled to a component is difficult to model. This uncertainty increases where components have geometrical features such as the grid pattern castellations that are often present on plasma facing components in fusion reactors. This project investigates the influence of various castellation patterns on the coupling characteristics of HIVE. It shows for example that as the grid density of castellations is increased, the applied heat flux increases from 4.5 MW/m2 to 6.93 MW/m2 for an input power of 30kW over 2 s, due to an improvement in the efficiency of the inductive coupling from 13.4% to 20.8%. Additional experimental factors affecting efficiency and homogeneity of heating are also discussed.

Progress in the production of the W7-X divertor target modules

The realization of the 19.6 m² highly heat loaded target area of the actively water-cooled divertor of Wendelstein 7-X (W7-X) requires the installation of 100 target modules distributed in ten discrete similar divertor units. A target module is water-cooled from manifolds to distribute the water equally between the target elements with flanged connections to the water supply system. A target element is made of a CuCr1Zr copper alloy heat sink armored with CFC NB31 tiles and designed to remove a stationary heat flux up to 10 MW/m². The production of the first 30 target modules has been successfully completed in the workshop of IPP-Garching. The quality of the finished target modules was systematically assessed as follows: visual inspections, measurement of the 3-D CFC target area, dynamic pressure tests, integral He leak testing under pressure at different conditions (20 °C, 4 MPa and 160 °C, 2.5 MPa) in a vacuum oven, high heat flux testing. Only 2 finished target modules did not pass the integral He leak testing and were successfully repaired.

The development of technology of Be/CuCrZr joining using induction brazing

The article describes the method of vacuum induction brazing of beryllium tiles covering the surface of the first wall of ITER. To prevent the formation of brittle intermetallic and preservation of the mechanical properties of the CuCrZr alloy, it leads to the need to minimize the brazing temperature and time of the brazing cycle. Designed according to the results of calculations and optimized in experiments, the inductor allows heating locally the beryllium tile/CuCrZr alloy region with required heating rate. With the aim to minimize brazing temperature this paper presents the result of selection of the optimal eutectic of STEMET 1101 brazing alloy thereby reducing the temperature of the brazing to 680 °C. To fix each of the beryllium tiles during the brazing cycle, a unique clamping tool was developed and applied. As a confirmation of the reliability of this method, along with the results of ultrasonic testing of brazed joints, the paper presents the successful results of high heat flux testing of the mock-ups throughout the full life-time of the ITER first wall panels (15,000 heat cycles).

Measures to overcome deep cracks of tungsten tiles in the ASDEX Upgrade divertor

A massive tungsten divertor, Div-III, was installed in ASDEX Upgrade (AUG) in 2014. Div-III is an adiabatically loaded component and consists of massive tungsten tiles clamped into their supporting structures. Before installing the new component, extensive studies, including Finite Element Modeling (FEM) and high heat flux tests in the test facility GLADIS, were carried out. After the first experimental campaigns the tile inspection reveals most of the tiles were cracked. The difference between the high heat flux tests and the AUG behavior was attributed to mechanical loads due to disruptions and/or the thermal load profile and history. The actions to understand the cracks comprise tests with the divertor manipulator, DIM-II, and FEM analysis of different target design options. DIM-II was used to test ‘split’ tiles, i.e. the deep crack is avoided by cutting a wide tile into two small tiles. FEM calculations were done to investigate the behavior of castellated targets with reduced tensile stress on top of the target and a clamping with a more elastic material, titanium instead of stainless steel. In addition more ductile tungsten heavy alloy was qualified for use in AUG. Based on this, a new set-up of tungsten tiles was installed in 2016 with different divertor configurations. After about 1000 plasma shots a thorough inspection of the tiles reveals that only 2 configuration out of 4 get rid of deep cracks in the tungsten tiles. Outcome of this inspection are here discussed and the final divertor setup is described.

Development of tungsten armored high heat flux plasma facing components for ITER like divertor application

The dome and reflector plate are the parts of divertor plasma facing components (PFCs) of ITER tokamak which are mainly aimed for the removal of heat load of maximum 5 MW/m2 in steady state condition. The dome is a curved tungsten armoured component and the reflector plate is a straight component. These components have multi-layered joints made of various materials such as tungsten (W), OFHC copper (Cu), copper alloy (CuCrZr) and stainless steel (SS316LN). Joining of such multi-layered joints is known to be problematic due to joining of several dissimilar materials. In this paper, we report the indigenous development of medium size dome and reflector plate via vacuum brazing route for ITER like tokamak application. In order to evaluate the performance of the dome against ITER-like scenarios (maximum heat flux removal of 5MW/m2), the dome has been successfully tested for 1000 number of steady-state thermal cycles at incident heat fluxes of 3.87 MW/m2 in the High Heat Flux Test Facility (HHFTF) at IPR. Subsequent testing of additional 200 thermal cycles was also done at incident heat flux of 6 MW/m2. During the High heat flux (HHF) tests, surface temperature of W tiles reached 640oC and the beam power was restricted at 6MW/m2 to limit the temperature below 450oC at the CuCrZr heat sink. Total 1200 steady-state thermal cycles have been completed. At 6 MW/m2, the absorbed heat flux was 4 MW/m2. Engineering analysis on the HHFT of the dome has been performed using Finite element method (FEM) and Computational Fluid Dynamics (CFD) to simulate and to correlate with the experimental data. Ultrasonic immersion technique – Non destructive testing (NDT) was used to inspect the brazed joint quality of the dome before and after the HHFT. The results of the experimental details, engineering analysis and methodology adopted to fabricate the medium size dome and reflector plate are presented here.

China’s technological achievements on ITER enhanced heat flux first wall in the pre-PA qualification

The Chinese Domestic Agency (DA) is procuring ITER Enhanced Heat Flux (EHF) First Wall (FW) panels which shall withstand a surface heat load up to 4.7 MW/m2 during ITER operation. Prior to the implementation of the ITER Procurement Arrangement (PA), several key technologies in manufacturing the EHF FW panel have been qualified. In the pre-PA task, ITER grade CN-G01 high-purity beryllium has been developed; its chemical composition, mechanical properties and grain size satisfy the ITER requirements. The strict quality control in the manufacturing of bimetallic plate was aimed at increasing the reliability the reliability of hypervapotron (HVT) heat sink for the EHF FW fingers, while a design improvement of the HVT cooling channel improved the thermal fatigue performance. Investigation showed that magnetron sputtering Ti/Cu coating on the beryllium tiles helps in reducing possible defects and results in better quality of Be-Cu hot isostatic press (HIP) bonding interfaces. As a result of such measures, China completed two EHF FW semi-prototypes with one successfully passing the high heat flux test (HHFT) at 4.7 MW/m2 for 7500 cycles and 5.9 MW/m2 for 1500 cycles. Neither damage to the semi-prototype nor off-normal surface temperature increase was observed, which brought the semi-prototype qualification campaign to a successfully conclusion.

High heat flux test results for a thermal break DEMO divertor target and subsequent design and manufacture development

This paper focuses on the development of the water-cooled divertor target concept known as Thermal Break, which was carried out in two phases. In Phase 1, six small scale mock-ups were fabricated and subjected to high heat flux (HHF) testing of up to 25 MW/m2 and thermal cycling of up to 500 cycles at 20 MW/m2. All six mock-ups survived the campaign and maintained 20 MW/m2 heat exhaust capability. Detailed examination of mock-ups was carried out to understand the damage mechanisms. One mock-up, which was tested beyond its design intent at 500 cycles, shows signs of progressive damage. Potential damage modes were identified and influenced subsequent Phase 2 mock-up design. Although there are signs of tungsten surface cracking, the predominant damage mode is not by “deep cracking” but substantial permanent deformation in the interlayer features.Therefore, in Phase 2 the manufacturing procedure was updated, the interlayer grooves were given stress-relieving radii which have significantly reduced the interlayer plastic strain range. Interlayer design parameters were selected following the use of response surface-based design search and optimization. Mock-ups of the Phase 2 design have been manufactured and HHF testing is planned within 2018.

Progress of the EU activities for the ITER Divertor Inner Vertical Target procurement

F4E undertook the qualification of so-called “additional suppliers” in order to enhance competition among the potential bidders and to secure the procurement of the ITER Divertor Inner Vertical Target. In order to assess the performances of W armoured plasma facing components under the conditions expected in the divertor target strike point region, a total of 36 W monoblock mock-ups were manufactured by Atmostat (F) by diffusion bonding, and by CNIM (F) and Research Instruments GmbH (D) by brazing, from which fifteen mock-ups were High Heat Flux (HHF) tested at IDTF (Efremov Institute Saint Petersburg, Russian Federation) electron beam test facility.The basic HHF testing program foresaw the performance of 5000 cycles at 10 MW/m2and 300+700 cycles at 20 MW/m2. The test results showed some significant improvements, in particular on the issues of W monoblocks macro-cracking and heat sink thermo-mechanical fatigue performances, identified during previous HHF tests campaigns. Some enhanced HHF thermal fatigue testing and critical heat flux experiments were also performed.

Repair processes of Wendelstein 7-X target modules

The highly loaded surface of the actively water-cooled divertor of Wendelstein 7-X (W7-X) is made of 100 individual target modules. In each target module, a set of target elements is water-cooled in parallel and fed by manifolds. A target element is made of a CuCr1Zr copper alloy heat sink, armored with CFC NB31 tiles. He leak testing under pressure in a vacuum oven was performed for each individual target element and module. One series target element of type 5S did not pass this test after high heat flux testing. The leakage was the result of the combination of a porosity concentration due to the overlap of CFC-CuCr1Zr welds in the centre part along the element and the reduced CuCr1Zr thickness at this location due to machined cut-out in the cooling channel. The repair process, which was successfully applied to all 307 target elements of type 5S, was sealing the slit between two CFC tiles by electron beam welding.Two target modules of type 7H did not pass the integral leak test. Leaks were localized by sniffing the weld seam between the target element connectors and manifold pipes. The one leak of the first module was successfully repaired by hand-welding. The positions of the two leaks in the other module did not allow such a repair process. Drilling apertures through the neighbouring manifolds allowed a direct access to the leaking seams from inside. The openings allowed a stable and precise installation of an inside orbital welding anode. The repaired target module successfully passed the He leak test in oven after HHF testing.

Materials development for new high heat-flux component mock-ups for DEMO

Material issues pose a significant challenge for future fusion reactors like DEMO. When using materials in a fusion environment a highly integrated approach is required. Damage resilience, power exhaust, as well as oxidation resistance during accidental air ingress are driving issues when deciding for new materials. Neutron induced effects e.g. transmutation adding to embrittlement are crucial to material performance. Here advanced materials such as tungsten fibre-reinforced tungsten Wf/W and fibre-reinforced copper Wf/Cu composites could allow the step towards a fusion reactor. Recent developments in the area Wf/W mark a possible path towards a component mock-up early enough for utilisation in DEMO. High heat-flux tests show that having short fibres at the exposed surface leads to their selective erosion and melting. Initial tests in the linear plasma device PSI-2 confirm this behaviour.

Ultrasonic analysis of tungsten monoblock divertor mock-ups after high heat flux test

In this paper, a comparison is presented among the ultrasonic measures of different W-monoblock mock-ups following high heat flux tests. The DEMO reference solution for the divertor targets (ITER-like) is compared with two other concepts, in which the interlayer between monoblocks and tube has been engineered to increase the performance of the component (Thermal Break Interlayer, Thin Graded Interlayer). Ultrasonic measurements have been performed at the depths of interest. The reflected signal amplitude and time of flight at defined depths are used to determine the shape and entity of the defect, when present. For each concept, the most frequently observed defects due to thermal loading are then identified. Preliminary post-mortem analysis confirms the results from the ultrasonic tests, proving the defect imaging capability of such non-destructive technique for advanced W-monoblock concepts.