Tungsten Monoblock Research Articles

High-heat-flux performance limit of tungsten monoblock targets: Impact on the armor materials and implications for power exhaust capacity

• Results of the extensive high-heat-flux tests of tungsten monoblock-type divertor target mock-ups are presented. • Impacts of various slow/short transient overloads relevant to European DEMO divertor on the armor materials and the joining interfaces are addressed. • Comprehensive high-quality data comprising in-situ IR thermography, ultrasonic reflectometry and electron microscopy (EBSD) are presented. • Implications of the novel test results for the power exhaust requirements are discussed. Development of a reliable high-heat-flux (HHF) technology is one of the crucial requirements of power exhaust strategy for a fusion reactor. The baseline HHF technology of the EU-DEMO, which inherited mostly the ITER technology, is based on the tungsten monoblock design and hot radial pressing joining technique. Thermal resilience and structural integrity under all off-normal transient events foreseen are essential prerequisites for validation of the technology towards the demonstration of full-scale prototype manufacture. Hence, the HHF performance of the baseline technology needs to be evaluated also in the transient overload regime. To this end, we conducted an extensive HHF testing campaign using small-scale test mock-ups of the tungsten monoblock target for two slow (10s) overloads at 20 MW/m 2 (up to 2000 pulses) and 25 MW/m 2 (1000 pulses), and for a short (0.4s) overload at 40 MW/m 2 (5000 pulses). Furthermore, excessive heat loads (32-37 MW/m 2 ) were applied beyond the armor melting event to test the structural stability at limit loads. IR thermography, ultrasonic inspection and electron microscopy (EBSD) delivered direct information and insight on the structural integrity and the impact on the armor microstructure. It was found that the tungsten monoblock target remained fully intact at least up to 1000 heating cycles at 20 MW/m 2 (10s) and survived 2000 cycles without any major failure. At 25 MW/m 2 (10s), the mock-ups remained nearly intact at least up to 500 heating cycles and survived 1000 cycles without critical failure. However, the armor surface showed substantial deformation roughening with the height of 1000 µm after 1000 cycles. At 40 MW/m 2 with short pulses (0.4s), the mock-ups remained fully intact without any serious damage at least up to 5000 cycles. The mock-ups even withstood the limit heat load of 32 MW/m 2 at least for a few pulses. Under excessive overloads (33-37 MW/m 2 ) above the critical incident heat flux, armor melting preceded any other potential failures (e.g. pipe rupture by coolant boiling).

Bidirectional reflectance distribution function of recrystallized tungsten mono-block exposed to cyclic heat loading

• BRDF of tungsten mono-blocks exposed to cyclic heat was measured. • Overall values of BRDF decreased with increase in cyclic heat loading. • Comparison between a modeled and measured BRDF was performed. • Additional diffused reflection model is necessary to a specular model to fit the measured BRDFs. The bidirectional reflectance distribution function (BRDF) of recrystallized tungsten mono-blocks exposed to cyclic heat loading for ITER divertor outer vertical target was measured using a goniophotometer at a wavelength of 473 nm. Laser microscope observations revealed that the surface roughness of the mono-block increased by an order of magnitude after cyclic heat loads of approximately 20 MWm −2 . Furthermore, the overall values of BRDF decreased with an increase in cyclic heat loading. A comparison between the modeled and measured BRDFs showed that an additional Lambertian component is necessary to a specular function (microfacet model or Gaussian function) to fit the measured BRDFs when the surface of the mono-blocks becomes rough.

Operating a full tungsten actively cooled tokamak: overview of WEST first phase of operation

WEST is an MA class superconducting, actively cooled, full tungsten (W) tokamak, designed to operate in long pulses up to 1000 s. In support of ITER operation and DEMO conceptual activities, key missions of WEST are: (i) qualification of high heat flux plasma-facing components in integrating both technological and physics aspects in relevant heat and particle exhaust conditions, particularly for the tungsten monoblocks foreseen in ITER divertor; (ii) integrated steady-state operation at high confinement, with a focus on power exhaust issues. During the phase 1 of operation (2017–2020), a set of actively cooled ITER-grade plasma facing unit prototypes was integrated into the inertially cooled W coated startup lower divertor. Up to 8.8 MW of RF power has been coupled to the plasma and divertor heat flux of up to 6 MW m−2 were reached. Long pulse operation was started, using the upper actively cooled divertor, with a discharge of about 1 min achieved. This paper gives an overview of the results achieved in phase 1. Perspectives for phase 2, operating with the full capability of the device with the complete ITER-grade actively cooled lower divertor, are also described.

Advances in the long-pulse steady-state high beta H-mode scenario with active controls of divertor heat and particle fluxes in EAST

Since the last IAEA-Fusion Energy Conference, the Experimental Advanced Superconducting Tokamak (EAST) research program has been, in support of ITER and CFETR, focused on development in terms of the long-pulse steady-state (fully noninductive) high beta H-mode scenario with active controls of the stationary and transient divertor heat and particle fluxes. The operational domain of the steady-state H-mode plasma scenario has been significantly extended with ITER-like tungsten mono-block divertor, plasma control and heating schemes. EAST has achieved several important milestones in the development of high β p H-mode scenario and its key physics and technologies. A 60 s-scale long-pulse steady-state high β p H-mode discharge with the major normalized plasma parameters similar to the designed performance of the CFETR 1 GW fusion power operation scenario has been successfully established and sustained by pure RF heating and current drive. Several feedback control schemes have been developed for a sustained detachment with good core confinement. This includes control of the total radiation power, target electron temperature, and particle flux measured using divertor Langmuir probes or a combination of the control of target electron temperature and AXUV radiation near the X point. The detachment feedback control schemes have been integrated with small-ELM regimes and high β p scenario via neon seeding, enabling a core and edge compatible integrated high-beta scenario applicable to long-pulse operations. ELM suppression has been achieved using various methods, including resonant magnetic perturbations and impurity seeding. Full suppression of ELMs by using n = 4 RMPs has been demonstrated for ITER for the first time in low input torque plasmas in EAST. EAST has been operated with helium to support the ITER research requirements for the first time. For a long-pulse, high bootstrap current fraction operation, a new lower tungsten divertor with active water-cooling has been installed, along with improvements in the heating and current drive capability.

Gas pressure casting technique for manufacturing of W/OFHC-Cu monoblock

The important engineering part of the divertor vertical target is the quality of joint between armor materials and heat sink. To reduce the joint interface stresses, a pure copper (Cu) interlayer was provided between the tungsten(W) armor and the CuCrZr cooling tube. Gas pressure casting method was developed to reduce the porosity and improve the mechanical properties of the Cu interlayer of W/Cu monoblock. In the vacuum casting sample without pressure, porosities were found in Cu. These porosities become defects at the interface between W and Cu and decrease the mechanical properties of the Cu interlayer. In the cooling process of casting, argon gas was pressurized into a porous tube inserted in the Cu and tungsten monoblock applying perpendicular force on the joint surface. This method was effective to remove the porosity in Cu generated during the cooling process and to increase the bonding strength between Cu and tungsten. The quality of bonding interface, depth of the Cu diffusion, and hardness of Cu were compared by SEM, EDS, and Vickers hardness measurement. In the samples made by gas pressure casting method, the porosity was decreased in Cu, no defects on the interface between tungsten and Cu, Cu was diffused into tungsten with a depth of 2∼3 μm and the hardness of Cu was increased about 15%.Small mock-ups were manufactured by hot radial pressing (HRP) with gas pressure casted W/Cu monoblock and CuCrZr cooling tube and underwent high heat flux (HHF) of 10 MW/m2 up to 5,000 cycles. The ultrasonic test (UT) results before and after the HHF test showed no defects on interfaces and was confirmed by observation of SEM on the cutting surface of the monoblock mock-up after the HHF test.

Fluid turbulence simulations of divertor heat load for ITER hybrid scenario using BOUT++

The BOUT++ six-field turbulence code is used to simulate the ITER 11.5 MA hybrid scenario and a brief comparison is made among ITER baseline, hybrid and steady-state operation (SSO) scenarios. Peeling–ballooning instabilities with different toroidal mode numbers dominate in different scenarios and consequently yield different types of ELMs. The energy loss fractions (ΔW ped/W ped) caused by unmitigated ELMs in the baseline and hybrid scenarios are large (∼2%) while the one in the SSO scenario is dramatically smaller (<1%), which are consistent with the features of type-I ELMs and grassy ELMs respectively. The intra ELM divertor heat flux width in the three scenarios given by the simulations is larger than the estimations for inter-ELM phase based on Goldston’s heuristic drift model. The toroidal gap edge melting limit of tungsten monoblocks of divertor targets imposes constraints on ELM energy loss, giving that the ELM energy loss fraction should be smaller than 0.4%, 1.0%, and 1.2% for ITER baseline, hybrid and SSO scenarios, correspondingly. The simulation shows that only the SSO scenario with grassy ELMs may satisfy the constraint.

Power exhaust concepts and divertor designs for Japanese and European DEMO fusion reactors

Concepts of the power exhaust and divertor design have been developed, with a high priority in the pre-conceptual design phase of the Japan–Europe broader approach DEMO design activity (BA DDA). Common critical issues are the large power exhaust and its fraction in the main plasma and divertor by the radiative cooling (P rad tot/P heat ⩾ 0.8). Different exhaust concepts in the main plasma and divertor have been developed for Japanese (JA) and European (EU) DEMOs. JA proposed a conventional closed divertor geometry to challenge large P sep/R p handling of 30–35 MW m−1 in order to maintain the radiation fraction in the main plasma at the ITER-level (f rad main = P rad main/P heat ∼ 0.4) and higher plasma performance. EU challenged both increasing f rad main to ∼0.65 and handling the ITER-level P sep/R p in the open divertor geometry. Power exhaust simulations have been performed by SONIC (JA) and SOLPS5.1 (EU) with corresponding P sep = 250–300 MW and 150–200 MW, respectively. Both results showed that large divertor radiation fraction (P rad div/P sep ⩾ 0.8) was required to reduce both peak q target (⩽10 MW m−2) and T e,i div. In addition, the JA divertor performance with EU-reference P sep of 150 MW showed benefit of the closed geometry to reduce the peak q target and T e,i div near the separatrix, and to produce the partial detachment. Integrated designs of the water cooled divertor target, cassette and coolant pipe routing have been developed in both EU and JA, based on the tungsten (W) monoblock concept with Cu-alloy pipe. For year-long operation, DEMO-specific risks such as radiation embrittlement of Cu-interlayers and Cu-alloy cooling pipe were recognized, and both foresee higher water temperature (130 °C–200 °C) compared to that for ITER. At the same time, several improved technologies of high heat flux components have been developed in EU, and different heat sink design, i.e. Cu-alloy cooling pipes for targets and RAFM steel ones for the baffle, dome and cassette, was proposed in JA. The two approaches provide important case-studies of the DEMO divertor, and will significantly contribute to both DEMO designs.

Contribution of leading edge shape to a damaging of castellated tungsten targets exposed to repetitive QSPA plasma loads

The castellated tungsten monoblock is considered one of the preferred reference designs of components faced to plasma in the divertor of a fusion reactor. The behavior of such components during the development of transient events (disruptions or/and of Edge Localized Modes (ELMs)) is still among the most critical issues for future thermonuclear devices. Experimental studies of castellated tungsten surface erosion have been performed within the powerful quasi-stationary plasma accelerator QSPA Kh-50 in conditions relevant to unmitigated ITER Type I ELMs. The surface energy load from the impacting hydrogen plasma streams was 0.9 MJ m−2 during 0.25 ms to achieve pronounced melting of the target surface. The studies of solid and liquid particle generation were performed at sequentially oblique plasma exposure of the different chosen edges of the castellated structure. It was determined that most of the W droplets are ejected from the sharp leading edge of the target due to the development of instabilities in the plasma-liquid metal interface. At the same time, the formation of solid dust particles is mainly attributed to the cracking during the surface cooling after the plasma pulse. Particle velocities typically achieve tens m s−1. Both the re-solidified droplets as well as dust particles have been collected on the special plate near the exposed surface. The maximal size of collected particles reached tens of micrometers. Such particles could either fly away with the next plasma pulses or mix with the material of the collecting plate in the course of repetitive plasma pulses. Thin (submicron) layers of re-deposited tungsten also formed on the surface of the collecting plate.

First feedback during series fabrication of ITER like divertor tungsten components for the WEST tokamak

Purpose of WEST is to test key technologies for future device like ITER. In particular the actively cooled plasma facing component technology, the so-called tungsten monoblock, will equip the lower divertor, allowing studies and analyses of the plasma wall interaction. 456 plasma-facing components have been manufactured during the last years, using hot isostatic pressing as main bonding process. The manufacturing of the components was not completely straightforward; however, it was successful and main requirements have been reached. The key manufacturing steps, as well as the associated controls are described.

Grain growth and damages induced by transient heat loads on W

During H-mode plasma in ITER, type-I Edge Localized Modes can occur naturally leading to transient heat load (<10 GW m−2) on the tungsten divertor. Such high heat fluxes might therefore induce annealing effects, such as recovery and recrystallization, or even lead to surface melting and deterioration (e.g. cracks) of the plasma-facing components. To ensure sufficient lifetime of tungsten monoblocks is thus mandatory to understand how W will evolve when exposed to such transient heat loads. We present here results on the study of grain growth and damage threshold of tungsten induced by transient heat loads (from 1 to 15 ms, <3 GW m−2). Using a high-power laser facility, we simulate ELMs and we submitted samples to various heating durations up to melting temperature. Laser heating is combined to several imaging techniques, like SEM and optical microscopy, to study grain growth and damages on W. We observed that the growth kinetics strongly depends on the repetition rate with a rapid grain growth for 100 Hz heat loads. We eventually built a finite element methods simulations, benchmarked against our experimental results, to evaluate temperature gradients and recrystallized fraction of the material when exposed to unmitigated ELMs.

In situ observation of tungsten plasma-facing components after the first phase of operation of the WEST tokamak

After three years of operation (2017–2019) and four experimental campaigns (C1 to C4), the first phase of operation of the WEST tokamak has now come to an end. During that period, WEST has operated with tungsten coated copper-based and carbon-based plasma facing components in order to comply with a full metallic environment. A set of components, called plasma facing units, using ITER tungsten monoblock technology was also installed onto one sector of the lower divertor to test their behaviour in a real tokamak environment. This paper gives an overview of the ageing behaviour of these components during in situ visual inspections. It is shown that tungsten coatings are in a good condition after 3 years of operation but are prone to delamination in the areas dominated by off-normal events such as runaway electrons or arcing. For this reason, the W coated carbon tiles of the outer limiter were replaced by bulk W tiles after the C3 campaign. As for the bulk tungsten ITER-like components, damage such as melting and cracking was mainly observed on the monoblock edges exposed to the parallel heat flux. The first observations show that the damage formed during C3 on the tungsten monoblocks did not evolve during C4, which leads to believe that it results from singular events rather than continuous heating due to steady state operation.

Multi-scale fracture probability analysis of tungsten monoblocks under fusion conditions

Plasma facing components inside future nuclear fusion reactors are subjected to a high heat load and intense irradiation conditions. Using advanced computational material models, several problems can be solved that reflect tungsten monoblocks under fusion relevant loading scenarios. This allows for the identification of the conditions under which material failure is probable. The material model and parameters are identified such that the mechanical behaviour is in accordance with the homogenized behaviour of a previously developed crystal plasticity model on the microscopic scale. The heterogeneous stress field that follows is analysed in order to assess the probability of material failure, which is typically reflected by unstable crack propagation. Since fracture is an inherently multi-scale problem, critical regions are analysed in detail by means of a representative volume element. The resulting analysis reveals that in case the stress relaxation in the monoblock under the applied static heat load is complete, the probability of unstable crack propagation can reach values close to 35%. Finally, the impact of prolonged neutron irradiation is simulated by means of a cluster dynamics model. Although irradiation drastically increases the brittleness of tungsten, its impact on the overall monoblock performance remains limited. • The performance of tungsten monoblocks under fusion conditions is investigated. • Failure is predicted by application of the weakest-link theory on two length scales. • Stress relaxation increases the risk of failure upon shut down of the heat source. • The impact of neutron irradiation on the failure probability is shown to be limited.

A numerical model for the recrystallization kinetics of tungsten monoblocks under cyclic heat loads

The recrystallization assisted microstructural changes during cyclic heat exposure are surmised detrimental to the thermal fatigue resistance of the tungsten monoblocks that cover the surface of the divertor components of a nuclear fusion reactor. A numerical framework to predict recrystallization in tungsten monoblocks during cyclic heat exposure is presented in this work. The framework is based on a thermal model coupled to a Johnson-Mehl-Avrami-Kolomogorov (JMAK) type recrystallization model. The influence of the initial (starting) microstructural state, i.e. the degree of deformation, on the extent of recrystallization, is considered through the activation energy for recrystallization. For a highly deformed microstructure, significantly more recrystallization is observed compared to self-diffusion driven recrystallization. The activation energy for recrystallization within the range 322 to 350 kJmol−1 during cyclic heat exposure of a monoblock is determined from the simulated recrystallization extent in relation to the experimental observations of differently processed monoblocks. The influence of the heating time per cycle for different activation energies on the recrystallization extent is also investigated. Instead of bulk recrystallization, only localized surface recrystallization is observed for a combination of a shorter heating time per cycle and a moderately deformed initial microstructure, which may contribute to improving the thermal fatigue resistance of the monoblocks.

CFD Analyses for the Upgrade Divertor System of KSTAR

The Korea Superconducting Tokamak Advanced Research (KSTAR) device, constructed in 2008, is a world-class superconducting tokamak fusion research device for the development of fusion energy. The expected heating power goal has been set to 12 MW by using an additional heating system, i.e., the second neutral beam injection (NBI) system NBI-2. As the heating power increases, resistance to high heat flux and cooling capacity at the divertor should be improved to exhaust power in the scrape-off-layer domain. Therefore, an upgrade of the divertor system for KSTAR was launched in 2019, and the upgrade divertor will be installed by 2022. The peak heat flux on the divertor target in steady-state operation is set to 10 MW/m2, and the ITER-like divertor type, the water-cooled tungsten monoblock, has been applied. The upgrade KSTAR divertor system comprises 64 cassette divertor modules. A divertor module consists of the inner target, the central target, the outer target, and the cassette body with supports to connect each part. In this study, thermal analyses were carried out to confirm the design’s thermal robustness for a whole divertor module. The temperature distribution and pressure drop were calculated by computational fluid dynamics analyses. Based on the response surface optimization method, the optimized tungsten monoblock design was derived. The optimized monoblock design showed that all materials, tungsten, Cu, and CuCrZr, comprising the divertor target, are operated within their allowable temperature windows. For the global divertor model applying the optimized monoblock design, steady-state and transient analyses were carried out for heat fluxes of 10 and 20 MW/m2. At 10 MW/m2, all composing materials were operated within the allowable temperature, while the maximum temperatures of tungsten, Cu, and CuCrZr exceeded the allowable temperature range of 20 MW/m2. However, the results were acceptable since the temperatures are sufficiently lower than the melting temperatures, and the slow transient case occurs quickly.

Conceptual design of a liquid-metal divertor for the European DEMO

Liquid metal (LM) divertors are considered for the European DEMO reactor, because they may offer improved performance compared to the tungsten monoblock concept. The goal of this work is to provide a concept design, and explore the limitations of liquid metal divertors. To this end, a set of design requirements was formulated in close collaboration with the EUROfusion Power Plant Physics and Technology team (responsible for the design of the EU-DEMO). Tin was chosen as the preferred liquid metal, because unacceptable Tritium retention issues arise when lithium is used in DEMO. A concept design was then chosen that consists of water cooled pipes that are square on the outside and round on the inside, a corrosion barrier, and a 3D-printed porous tungsten armor layer filled with liquid tin. The porous armor layer acts as a Capillary Porous System (CPS). The design was analyzed using thermo-mechanical FEM simulations for various armor thicknesses and heat sink materials: Densimet, W/Cu composites, and CuCrZr. The highest heat loading capability achieved is 26.5 MW/m2 in steady state (18.9 MW/m2 when taking into account a safety margin of 1.4). This is achieved using a CuCrZr pipe, with a 1.9 mm thick armor. When increasing the armor layer to 3 mm thick, more than 80 MW/m2 can be withstood during slow transients thanks to vapor shielding, but at the same time the steady-state capability is reduced to 18 MW/m2. Resilience against disruptions cannot yet be proven, but is deemed within the realm of possibility based on estimates regarding the behavior of vapor shielding. This should be further investigated. Overall, the concept is considered a significant improvement compared to the original specifications (which are also the specifications to the tungsten monoblocks: 10 MW/m2 in steady state, and ∼20 MW/m2 during slow transients). Moreover, the possibility of withstanding disruptions is regarded as a potentially major improvement.

Evaluation of tensile and elastic properties of W/Cu cold-spray coatings for application to the FGM DEMO divertor concept

The ITER based conceptual design of EU DEMO divertor envisages tungsten monoblocks as plasma-facing material joined to a CuCrZr cooling tube with a thick Cu-OHFC interlayer. Since divertor configuration will define the power exhaust capabilities of the reactor, additional concepts have been developed within the EUROfusion WPDIV project. In this regard, FGM (Functionally Graded Material) concept is one of the proposed alternatives. The primary motivation of this concept is to replace the thick copper interlayer with a W/Cu compositionally graded thin film for armour-to-pipe joining with the aim of minimizing the stresses at the interfaces while assuring the structural integrity of the component. This FGM interlayer will be composed of stacked elementary layers manufactured by Cold-Spray with the following four target compositions: 100 vol.%Cu, 25 vol.%W + 75 vol.%Cu, 50 vol.%W + 50 vol.%Cu, 75 vol.%W + 25 vol.%Cu. The elastic modulus of these coatings is an especially important materials parameter since it has a strong influence on both FGM fracture and stress concentration at the interface. Therefore, it is of fundamental importance to determine the elastic modulus of the as-deposited coatings. In this work, the mechanical and microstructural characterisation of each elementary layer is presented, and different methods for the evaluation of their elastic modulus have been evaluated, and their agreement has been assessed. Tensile tests were conducted on the free-standing coatings to understand the strengthening and deformation mechanisms of each composition. The ultimate tensile strength increased proportionally with the W content, showing 50% higher tensile strength on the addition of less than 40 vol.%W since crack arresting tungsten particles were observed as strengthening mechanism.

Impact of W monoblock damage on EAST operations in recent campaigns

The upper divertor of EAST has been upgraded to ITER-like tungsten plasma-facing components since 2014. The impact of tungsten monoblock damage on plasma operations in recent EAST campaigns was studied. Various damage on the divertor tiles such as melting, cracking and arcing was found. Statistics analysis on the disruptivity due to W monoblock damage have been performed. W impurity produced from W divertor damage becomes the main impurity that leads to disruption during 2018, 2019a and 2019b campaigns compared to other impurities. With increasing auxiliary heating power, the W divertor damage becomes more and more serious. The current divertor damage is still tolerable for the overall EAST operations. Such studies can provide good references for the future fusion devices with W monoblocks, such as ITER.

Design and integration of femtosecond Fiber Bragg gratings temperature probes inside actively cooled ITER-like plasma-facing components

Measuring the temperature in plasma-facing components (PFCs) provides information both on plasma parameters in the divertor region and on the thermal stress experienced by PFCs. Fiber Bragg gratings (FBGs) are interesting candidates for this application because they are immune to electromagnetic interferences and their ability to be multiplexed allows an extended spatial coverage. Four fibers, each of them including eleven regenerated Bragg gratings, have been embedded in tungsten-coated graphite components and operated up to their signal-collapsing limit at 800 °C. Extending the measurement range towards higher temperatures increases the sensitivity to plasma parameters and allows withstanding higher energy experiments. To overcome thermal limitations, the system is up-graded using femtosecond laser inscribed fibers. In addition to their outstanding thermal stability, femtosecond FBGs benefit from higher signal-to-noise ratios than regenerated FBGs. The paper addresses femtosecond FBGs design and issues relative to their integration inside the actively cooled ITER-like PFCs of the WEST tokamak. The period and length of the gratings are designed to increase the number of measurement spots to fourteen gratings per fiber, regularly distributed over 17 cm, while ensuring robust detection even with strong thermal gradients (no overlapping or deformation of Bragg peaks). The system operates up to 1200 °C with gradients reaching 200 °C/mm perpendicularly and 40 °C/mm in parallel to the fiber. FBGs are inserted in actively cooled ITER-like PFCs through a 2.5 mm deep lateral groove localized at 5 mm beneath the top of bulk tungsten monoblocks. A PFC mock-up machined with a groove has been tested under HHF facility to assess the effect of the groove on monoblocks thermal behavior. The test demonstrates that machined monoblocks behave as expected from simulation and can withstand 20 MW/m2 heat flux (i.e. 1200 °C in the fiber) with 20 % overheating as compared to intact monoblocks.

Analysis of hardness and microstructural changes in Tungsten mono-blocks exposed to high heat flux at 10 MW/m2

In this study, we investigated the material properties of tungsten mono-blocks undergone high heat flux test. The ITER-like tungsten mono-block specimens were fabricated by two kinds of joining methods. The tungsten-Cu interlayer joint was made by gas pressure casting and Cu interlayer-CuCrZr cooling tube were joined by hot radial pressing (HRP) or brazing method. To evaluate the integrity of the fabricated mono-blocks, high heat flux tests were done on the small mockups under cyclic heat load of 10 MW/m2 up to 5,000 cycles. We measured the microstructure, hardness, and surface state of the tungsten armors before and after the heat flux exposure experiment. The Vickers hardness profile measured from the surface down to the Cu interlayer showed a noticeable reduction around 200 μm ~ 500 μm depth after high heat flux test compared to that before the test depending on the bonding quality of mockup specimens. At depth beyond 800 μm, the Vickers hardness did not show any notable difference before and after the high heat flux test, which indicates the material property of tungsten is not so much affected in this region at the above heat flux condition. The microstructure and grain size distribution was measured and analyzed by SEM and EBSD. The correlation between the microstructure and hardness profile, and the impacts of the mockup performance under HHF test to the integrity of tungsten armors are analyzed.

Performance evaluation of tungsten for ITER divertor toward mass production

For the series production of the divertor outer vertical target (OVT) of the ITER, approximately 200,000 tungsten (W) monoblocks as plasma-facing material are planned to be manufactured. However, thus far, the variation in the properties of the W monoblocks has not been investigated. In this study, the chemical impurity level, material properties, and response to cyclic heat loading (i.e., durability and surface deformation) of W monoblocks manufactured using W plates supplied by A.L.M.T. Corp., Japan (ALMT) and Nidec Sankyo CMI Corp., Japan (NSCM) were investigated. On examining the manufactured lots (four from ALMT and two from NSCM), it was concluded that the variation in the material properties might have been caused by variations in the impurity level of the raw powder, contamination during manufacturing, and changes in the processing parameters. The results of the high heat flux test indicated that all the examined lots met the acceptance criteria. Hence, the variations observed in the chemical impurity level and material properties (e.g., density, Vickers hardness, and grain size) of the W plates supplied by ALMT and NSCM were acceptable. Therefore, W from the two suppliers will be employed in the full-scale plasma-facing unit prototype, which will be used for the final qualification of the material and OVT manufacturing technology.