High Heat Flux Tests Research Articles

Inside high-temperature ablation phenomenon and analysis of mechanism of a flat-type W–ZrC/Cu high heat flux mock-up

As a plasma-facing material (PFM), W–ZrC is a promising candidate for the divertor in future fusion reactors. To verify the feasibility of W–ZrC, a high heat flux (HHF) test was conducted on a comprehensive experimental platform. A rare internal ablation failure mode was observed during the experiment. Considering the internal ablation phenomenon during the test, the measured data of the test platform were analyzed to predict an ablation time of 12 s. Based on the shape and size of the ablation pit, the cause and mechanism of ablation were determined using thermal simulation analysis. It can be concluded that a sudden and significant increase in heat flux combined with reduced water flow can cause thermal ablation to occur within the heat sink. Furthermore, dimensions of the ablation pit are related to the area in which the heat flux is imposed. In sum, internal ablation can have serious consequences for the divertor.

Development and manufacturing of beryllium-armoring ITER enhanced heat flux FW toward series production in China

ITER’s enhanced heat flux (EHF) first wall (FW) provides thermal shielding to the other components behind it and limits its plasma boundary, consequently bearing high surface heat loads up to 4.7 MW m−2. China developed the EHF FW technologies in steady stages by making small mock-ups, semi-prototypes and a full-scale prototype (FSP), working toward series production at the Southwestern Institute of Physics. All the key technologies for bimetallic diffusion bonding, welding and assembly have been fully qualified. The Be armor tile size effects on thermal-fatigue performance have been evaluated using a pulsed high heat flux test with an EMS-400 electron beam facility. It was found that both the Be/CuCrZr bonding and its thermal-fatigue life are highly dependent on the tile size, and 12 × 12 mm2 Be tiles are acceptable for the required performance, either as individual tiles or in castellation by cutting larger ones. Small-scale mock-ups with such Be tiles survived 16 000 thermal cycles at 4.7 MW m−2, while EHF FW fingers with 16 × 16 mm2 Be tiles demonstrated unstable performance. In contrast, only BE tiles larger than 24 × 12 mm2 showed reliable diffusion bonding with the CuCrZr heat sink by hot isostatic pressing and consequently should be castellated into smaller ones. The manufacturing of the FSP finger goes through complex thermal cycles, using CuCrZr alloys in age-strengthening plus cold-rolling states, enabling it to maintain its tensile strength to the level of 285 MPa and its mean grain size less than 100 µm. The pipe welding for finger pairing and its assembly with the central beam are demonstrated, including enabling welding to be carried out twice for repair in a narrow space. The FSP showed good finger alignment, dimensional control and reliable vacuum tightness without any blockage of cooling channels.

Numerical study and experimental validation of heat transfer capacity of flat-type divertor for fusion reactor

The divertor must simultaneously withstand unprecedented high heat fluxes of up to 20 MW/m2 and high-energy neutron irradiation of 14-MeV during fusion reactor operation. Accordingly, it is necessary to simultaneously meet the excellent heat dissipation capacity of the divertor and maintain good material performance in harsh neutron irradiation environments. The flat-type divertor demonstrates better heat transfer performance compared to monoblock divertor. Nevertheless, under the condition of a heat flux of 20 MW/m2, the heat transfer and thermal fatigue performance of flat-type divertor using advanced materials are still unidentified. In this study, we conducted a thorough analysis of the heat transfer capabilities and thermal fatigue characteristics of potassium-doped tungsten (KW)/Cu/Oxide dispersion strengthened copper (ODS-Cu)/reduced activation ferritic/martensitic (RAFM) divertor mockup adopts hypervaportron (HV) structure by numerical simulations combined with experiments. The numerical results indicate that the flat-type KW/Cu/ODS-Cu/RAFM divertor mockup expresses excellent heat transfer capacity. The contact area between the edge of the fins and the bottom surface of the ODS-Cu heat sink induces a significant increase in water flow velocity to ∼15 m/s. The peak temperature of loaded KW surface is only ∼985 °C under the flow rate of 5 t/h and inlet temperature of 20 °C. During the high heat flux tests, the prepared flat-type divertor mockup successful endured 1000 cycles of 20 MW/m2 with the peak temperature of 883 °C, and the surface temperature experienced a fluctuation of 2.4 % during the thermal fatigue tests. This study can provide a strong data reference and technical support for the development of fusion reactors, and is of great significance in advancing the commercialization of fusion energy.

Fabrication of NBI Ion Source Back Plate and its High Heat Flux Experiment

The back plate is an important component of the ion source because of its multiple roles including heat load removal during beam operation. The main components of the back plate are (1) a Type 304L stainless steel (SS304L) magnet positioning plate that holds samarium cobalt permanent magnets required for the confinement of ion source plasma, (2) an oxygen-free electronic copper cooling plate with 35 inner and 8 outer cooling channel grooves (each of which is 4 × 1.8 mm2) that is vacuum brazed with a SS304L magnet positioning plate, and (3) a SS304L magnet cover plate. In this paper, the back plate is successfully fabricated, and a high heat flux experiment is done at the High Heat Flux Test Facility Center with an electron beam power of 200 kW for 458 s. The uniform incident heat flux is 2.5 MW/m2. Demineralized water at 34°C is supplied at the rate of 1 kg/s to the cooling plate at inlet pressure of 8.2 bars to remove the high heat load. The surface temperature of the copper plate is measured by an infrared camera, and three temperature regions are observed. The measured average surface temperature of the cooling plate is ~152°C. The bulk water temperature rise ΔTw is ~39.42°C. The estimated absorbed heat flux is ~2.04 MW/m2, and the heat absorption coefficient is 81.6%. The measured leak rate after the heat flux test is 1.6 × 10−8 mbars∙L/s. These High Heat Flux Test experimental results will be useful to study the thermomechanical performance of the back plate and to understand the effect of increasing the beam pulse length.

Manufacturing trials of PFCs with low thermal conductivity features for limiters

In future demonstration fusion power plants, plasma facing components will face high steady state and ultra-high transient heat loads from the plasma. Introducing low thermal conductivity features to the component can reduce the peak temperatures seen by the coolant pipe during transient loads, but at the trade-off of a loss of steady-state performance. Producing low-conductivity tungsten by additive methods has been investigated elsewhere, so this trial investigates production through subtractive means i.e., conventional milling. This method preserves the original temperature limits and majority of the microstructure and is much higher Technology Readiness Level (TRL) so requires less development to reach series manufacture. In this trial, diamond drilling produced holes in monoblocks down to 0.6 mm and ligaments down to 0.7 mm, showing that this method is capable of producing small features in a controlled way. With the geometric design of lattice used in this trial, these samples achieved thermal conductivity reductions of 15% to 40%. One monoblock assembly was tested up to 1100 °C for 500 cycles in the Heating by Induction to Verify Extremes (HIVE) High Heat Flux test rig and showed no signs of ligament cracking or thermal degradation. Electromagnetic modelling has been validated against the experimental results, so future designs can be accurately modelled ahead of testing in HIVE.

Development of brazing technologies for bonding of graphite and CuCrZr joints applying for the first wall of the EAST device

Bonding between plasma-facing materials (PFMs) and heat sinks is an important factor for the performance of plasma-facing components (PFCs) in the tokamak device; in some specific applications, graphite is adopted as the plasma-facing material and CuCrZr as the heatsink. Bonding of graphite and CuCrZr is a critical and difficult process due to the poor wettability of graphite and wide gap of material properties between graphite and CuCrZr, to which should be paid more attention. Herein, brazing processes were developed to achieve the bonding of CuCrZr and graphite, and a layer of oxygen-free copper (OFC) was placed between graphite tiles and CuCrZr heat sinks as a stress-buffer layer. Three types of brazing fillers including Ti powder with AgCu foil, AgCuTi paste and AgCuTi foil were used to achieve the bonding of CuCrZr and graphite. Samples were tested by scanning electron microscopy (SEM), energy dispersive spectrometry (EDS) and shear strength experiment; the results showed that these three types of fillers can achieve the bonding of CuCrZr and graphite. In addition, one-step brazing and multi-step brazing were both developed to achieve the bonding of graphite/OFC/CuCrZr, and the mock-ups were manufactured and endured cyclic high heat loads in the high heat flux (HHF) tests. In conclusion, the simplified one step brazing process including types of fillers, brazing temperature and brazing sequence was determined, which provides a promising premise method for the heat transfer structure design of PFCs.

Electron beam profile measurement using enhanced dual-techniques in high heat flux test facility at Institute for Plasma Research.

During operation of an ITER-like plasma fusion device, Plasma Facing Components (PFCs) of a divertor are subjected to steady-state and transient heat loads of up to 20 MW/m2. At High Heat Flux Test Facility (HHFTF) at Institute for Plasma Research, PFCs are subjected to such heat flux scenarios using 200 kW electron beam (EB). A heat flux distribution on PFCs is significantly influenced by the EB profile and scanning technique. A novel dual-technique approach is used to measure EB cross-sectional profiles in HHFTF. A single setup is designed, fabricated, and installed in HHFTF consisting of (1) the knife-edge technique involving graphite and tungsten rectangular tiles bonded on a copper block adjacent to each other to form a sharp edge and (2) the wire scanner technique wherein tungsten wires are placed at equal distances. Current collected by the setup is measured as EB is rastered across the wire and tile boundaries. Current measured as a function of EB position is used to calculate the EB profile using a customized LabVIEW software application. Experiments are conducted by varying EB power from 28 to 200 kW and correspondingly the full width at half maximum of the EB profile grows from 5.30 to 16.04mm, as observed by both these techniques. The COMSOL Multiphysics simulation of the wire scanner technique is performed to verify the experimental results. X-ray and infra-red imaging diagnostics of HHFTF are also used to measure the EB profile in a unique and first-of-its-kind way, and they are found to agree well with the dual-technique measurements.

High heat flux components with beryllium armour: From the small-scale mock-ups to the full-scale prototype of the ITER first wall panel

The Russian Federation shall supply 179 enhanced heat flux First Wall Panels (FWP) to the ITER construction site. These panels have about 30 types and consist of a steel supporting structure (FW Beam) with multilayer plasma-facing units (PFU) attached onto it. To start the production of the panels, the manufacturing readiness must be confirmed by the ITER Organization (IO). The JSC “NIIEFA” responsible for manufacturing of a part of the components, final assembly and testing of the panels has come a long way from the experimental work on small-scale mock-ups, through the development of the design and key manufacturing technologies to the production and testing of the full-scale prototype (FSP) which is the final step before the confirmation for the manufacture. The manufactured FSP has successfully passed most of the acceptance tests including hydraulic pressure and leak tightness tests. The high heat flux tests have been successfully carried out on the PFUs with beryllium armour. In parallel to the tests of the beryllium prototype, the tungsten armour has been analyzed and tested which is due to the IO's intention to change the armour material.

Interface cracking damage of W/Cu monoblocks as divertor target after high-cyclic plasma discharges in EAST

W/Cu monoblocks will be employed in ITER as divertor targets for high heat loads. Experimental results of W/Cu monoblocks in high heat flux (HHF) tests have demonstrated their endurance on 5000 cycles at 10 MW/m2 and occasionally 300 cycles at 20 MW/m2. However, plasma-facing units (PFUs) serve in a harsh and complex environment in tokamaks and whether W/Cu monoblocks can show the same excellent properties in tokamaks needs to be further tested. Experimentally evaluating the evolution of ITER-like PFUs is one of the main missions for the EAST tokamak. The ITER-like actively cooled W/Cu monoblocks installed in the EAST lower divertor in 2021 exhibited abnormal behavior after 13,346 plasma discharges. The abnormal temperature evolution of the component was observed via infrared cameras (IR). Subsequent non-destructive testing (NDT) indicated that it might primarily arise from interface damage and scanning electron microscopy (SEM) confirmed that there were obvious cracks at the interface. The cracking center was also found above the interface. Simulation of the structural characteristics of W/Cu monoblocks under heat load in EAST can obtain temperature distribution and stress–strain distribution, which can be used to analyze the causes of material interface cracks. These findings unveil the phenomenon and mechanism of interface damage in W/Cu monoblocks after enduring over 10,000 plasma discharges in EAST, providing valuable insights for the future long-term application of W/Cu monoblocks in ITER and fusion reactors.

Blanket test facility for mockup testing on water cooled ceramic breeder blanket system

Intensive research and development activities have been conducted for designing a water-cooled ceramic breeder (WCCB) blanket system. In this system 325 °C/15.5 MPa of pressurized water removes heat injected and accumulated in blanket containers made of a reduced activation ferritic/martensitic steel, F82H. The container has lithium ceramics and beryllium pebbles as tritium breeder and neutron multiplier, respectively. In order to demonstrate performance of the system, especially for high-temperature pressurized water coolant, experimental apparatus have been installed in Rokkasho Fusion Institute, National Institutes for Quantum Science and Technology, Japan. These apparatuses enable to conduct the following experiment on full-scale mockups: 1) high heat flux (HHF) test to demonstrate heat removal of surface heat flux by electron beam gun simulating plasma, 2) flow assisted corrosion, 3) coolant ingress into dummy pebble beds, and 4) beryllium-water reaction between the coolant and the pebbles. In this paper, overviews, specifications, early results of these apparatus are reported.

Experimental investigation on heat transfer characteristics of a novel hypervapotron channel for divertor target in nuclear fusion reactors

In the further development of nuclear fusion energy, managing the high heat flux on a divertor target remains a significant issue. To address this, the development of innovative cooling channels capable of efficiently managing high heat loads is essential. Among various cooling techniques, the hypervapotron (HV) method stands out for its exceptional heat transfer capabilities. This study undertakes an in-depth analysis of a novel HV cooling channel, evaluating its heat transfer performance via an electron beam high heat flux testing system. The focus was on subcooled flow boiling, observing that the novel HV channel displayed varied flow patterns with increasing heat flux. Additionally, a comparative investigation was conducted to assess the heat transfer efficiency of this novel HV channel against traditional ITER-like HV channel and smooth channel. The analysis incorporated various system parameters, such as mass flow rate and system pressure, to understand their effects on heat transfer. To effectively predict the heat transfer performance of the novel HV channel, efforts have been made to evaluate existing and develop new heat transfer correlations for the channel. Furthermore, Artificial Neural Networks (ANNs) have also been utilized as a crucial approach in this predictive endeavor. The outcomes of this study hold significant implications for understanding and optimizing the design of cooling channels for divertor target in nuclear fusion reactors.

Design and construction of a helium cooling experimental loop

The helium cooling tritium breeding blanket and helium cooling divertor, which produce tritium or heat, are important components in Chinese Fusion Engineering Test Reactor (CFETR) design. The helium cooling system, which function is similar to the primary loop of Pressurized Water Reactor (PWR), is designed to extract energy from blankets and divertors. These helium cooling components have many complex internal flow channels to extract high thermal load. They also have to withstand electromagnetic force, thermal stress and coolant pressure, etc. Therefore, out-of-pile experiments must be carried out as far as possible to verify the design. Under CFETR project support, a helium cooling experiment loop is built to carry out the thermo-hydraulic experiments for these components and to accumulate experience of helium cooling system. According to the cooling requirements of blanket, the loop test section operating parameters are flow rate 2.5kg/s, temperature 550°C and pressure 12MPa. During preliminary commissioning, these parameters have been reached. A high heat flux testing facility will connect to this loop for future experiments.

Consideration of tungsten recrystallization in plasma facing components design

Tungsten recrystallization leads to a modification of material properties including a reduction in hardness, strength and thermal shock resistance. The recrystallization assisted material degradation compared to the as-received state has made it very undesirable as demonstrated in high heat flux testing. Consequently, tungsten recrystallization temperature has been widely used as the upper limit of material surface temperature under steady state loading conditions. However, this study shows that recrystallization is inevitable where steady state or transient loads are high enough. The efforts to retard recrystallization will only reduce the recrystallized depth but cannot completely avoid it. While it is demonstrated the material will be weakened by recrystallization, the recrystallized depth plays a relatively small role in deciding the fatigue lifetime once it is recrystallized. Therefore, it is recommended that the geometry and fracture toughness of the design is optimised to reduce crack growth into the component. The design criterion against tungsten recrystallization under steady state loading conditions has also been proposed for engineering design stages.

Design of the CFC composite brazed HFS first wall on the NBI shine-through area of the EAST device

The high field side (HFS) first wall is the inner wall nearest to the vertical center axis in tokamak. To avoid high Z impurity sputtering due to plasma interaction and neutral beam injector (NBI) radiation, graphite is adopted as the plasma facing material (PFM) of the HFS first wall on the NBI shine-through area (NBISTA). In the experimental advanced superconducting tokamak (EAST), graphite tiles are bolted to the CuCrZr heat sinks by screw on the NBISTA, but cracks, fractures and ablations were observed on the graphite tiles after experimental campaigns because of insufficient heat removal capability. Thus, a new design of the HFS first wall on the NBISTA was proposed by adopting the carbon fiber composite (CFC) and brazing technology. In the new design, CFC tiles were brazed to the CuCrZr heat sink, and the oxygen-free copper (OFC) sheet was adopted between CFC tiles and CuCrZr heat sinks as stress-buffer layer. To avoid potential brazing failures or CFC tiles detachments, a modular pattern active cooling panels was proposed for the HFS first wall on the NBISTA. Steady heat transfer analysis of those panels was carried out and results show that the heat removal capacity satisfied the design requirements. Meanwhile, feasibility of the CFC brazing process was verified by nondestructive testing (NDT), mechanical property testing and high heat flux (HHF) testing, and testing results indicate that the CFC/OFC/CuCrZr bonding processed by one-step brazing had good mechanical property and heat removal capability.

Thermal and structural analysis of JT-60SA actively cooled divertor target submitted to high heat flux

JT-60SA is a joint international fusion experiment being built and operated in Japan under the framework of the Broader Approach Agreement and Japanese National Fusion Programme, aiming at an early realization of fusion energy by conducting supportive and complementary work for the ITER project, all towards supporting the basis for DEMO. In this context, a collaboration has been established via EUROfusion between Fusion for Energy and CEA to develop actively cooled targets of the JT-60SA divertor.The first Actively Cooled Divertor (ACD) target, planned to be operated in 2029, has graphite as armour material and TZM as heat sink material. To be DEMO relevant, ACD target with tungsten armour is planned to be operated in 2033, with either CuCrZr or TZM as heat sink material.This paper focusses on the choice of the design (material, dimensions) of these targets. Results of thermal and structural analysis of these targets are presented. The calculations, carried out in the environment of Ansys 2021R2 code, have taken into account the heat loading under plasma operation (with a peak heat flux of 10 MW/m2) and the electromagnetic loads from a representative plasma disruption (VDE 30 ms). In order to check the compliance with nuclear requirements, RCC-MRx 2015 has been used for the mechanical assessment of metallic components, while for the graphite the comparison between the maximum stress and the yield stress has been considered. The performed analyses show that the thermal load on target during the plasma heat loading drives target designs due to differential expansion at the interface between the heat sink and the flat tiles.Additionally, on-going assessments on these divertor targets are presented, focussing on experimental High Heat Flux testing, as well as computational thermal-hydraulic studies on enhanced target configurations.

High Heat Flux Tests in Support of the 3-D Computational Modeling of Melting for the EU-DEMO First Wall Limiters

High Heat Flux Tests in Support of the 3-D Computational Modeling of Melting for the EU-DEMO First Wall Limiters

High Heat Flux Testing of Brazed W/CuCrZr Monoblock in High Heat Flux Test Facility

The Institute for Plasma Research (IPR) is involved in the development of various technologies for fabrication and testing/qualification of tungsten (W)–based plasma-facing components for ITER-like applications. As a part of these activities, a small-scale W/Cu/CuCrZr monoblock test mock-up is prepared by using a vacuum brazing process. Further, thermal fatigue testing of this test mock-up is performed using the High Heat Flux Test Facility at IPR under an absorbed heat flux of ~9.5 MW/m2 for 1200 thermal cycles. Thermal-hydraulic analysis and computational fluid dynamics (CFD) simulations are also performed to understand the heat flux testing scenario. Nondestructive testing of the test mock-up after thermal cyclic heat flux testing indicates that there are no defects in the test mock-up except for minor cracks on the surface of W tiles. The results of thermal cyclic testing, finite element analysis, CFD analysis, and postanalysis of the High Heat Flux test are presented in the paper.

Benchmarking by high heat flux testing of W-steel joining technologies

For a future commercial fusion reactor, the joining of tungsten and steel will be of vital importance, covering the main part of the plasma facing area. However, the large difference, of more than a factor of 2, in the coefficient of thermal expansion (CTE) of W and steel results in high thermal stresses at their interface. The cyclic nature of the operation can cause fatigue effects and could result in a premature failure of the joint.One possible solution is the insertion of a functionally graded material (FGM), with varying the CTE, as an interlayer between tungsten and steel, which could reduce these stresses. In this study, two processes, atmospheric plasma spraying (APS) and spark plasma sintering (SPS), are utilized to manufacture such FGMs. The gradation was accomplished by using two or three layers with a thickness of 0.5 mm each.Another principle is the insertion of a ductile metal interlayer, which reduces the stress by plastic deformation. Vanadium and titanium foils of varying thickness were chosen, as both have a CTE in between W and steel and V forms a solid solution with W and Fe. These and a direct W-steel joint as baseline reference were made by current-assisted diffusion bonding. All samples consist of 3 mm thick W and steel tiles allowing a direct comparison of the different technologies.An efficient high heat flux benchmark test procedure was developed and performed to investigate and compare the potential of the different joining technologies. For this, the complete stacks were brazed on actively cooled copper cooling modules and tested with high stationary heat loads of up to 5 MW/m2 with 200 cycles at each level in the JUDITH 2 facility. Detailed thermal analysis including comparison with prediction based on FEM simulation are presented to understand the cause of the failure and track the degradation. This study allows to help focusing the further development of W-steel joining technologies.

A transient heat flux sensor for PbTe thin films based on transverse Seebeck effect

Among thermoelectric materials with thermoelectric effect, lead telluride (PbTe) is widely used because of its high performance and chemical stability in the medium temperature region. In this study, PbTe was creatively employed to develop a transient thin films heat flux sensor (THFS) using magnetron sputtering technique based on the transverse Seebeck effect. After static and dynamic calibration, the rise time of the THFS is 35 μs and the sensitivity is up to 7.9 μV(kW·m−2)−1, it can be measured in transient high heat flux testing environments without the need for a signal amplifier. In the experiment of measuring transient heat flux with explosion driven shock tube, the THFS has the advantages of high resolution and high dynamic response, which provides scientific basis for the study of explosion thermal damage effect and has important significance.

Development and Application of Vibration Diagnostics for Condition Monitoring in High Heat Flux Test Facility at IPR

Development and Application of Vibration Diagnostics for Condition Monitoring in High Heat Flux Test Facility at IPR