ITER-grade Tungsten Research Articles

In-situ TEM investigation of recovery mechanisms in ion-irradiated ITER-grade tungsten

The annealing of irradiation defects in tungsten was investigated under in-situ observations conducted using transmission electron microscopy (TEM). ITER-specification tungsten was exposed to heavy-ion irradiation by 10 MeV Au+ ions to 0.5 and 1 dpa at room temperature to generate defects. Subsequently, the thermal stability and recovery mechanisms of the generated defects are investigated by in-situ TEM at 600 °C, 800 °C, 900 °C and 1000 °C. The obtained TEM data are used to visualize and quantify the defect characteristics such as loops and voids as well as for observing the evolution of the microstructure and defect diffusion/recovery mechanisms. Focus is made on specific spatial phenomena such as loop-loop interaction, loop-dislocation and loop-grain boundary decoration, loop pinning, entanglement, formation of raft and transformation of loops into a dislocation network. The obtained results are in line with earlier obtained data for pure recrystallized tungsten, still, present results provide important information such as radiation-induced defect characteristics and recovery mechanisms under high temperature for the ITER-specification tungsten.

Amorphous and anisotropic surface relief formation in tungsten under repeated high-flux hydrogen plasma loads

Facing extreme plasma loads, the structural integrity of the tungsten divertor is crucial for ITER, an engineering marvel in nuclear fusion reactors, to achieve its fusion performance targets. Induced by repeated transient heating from plasmas, the thermal fatigue damage of tungsten–typically accompanied by the formation of surface relief–has been identified as a critical issue but an in-depth understanding is still lacking. Here, we report the formation of amorphous and anisotropic surface relief on ITER-grade tungsten surfaces under ITER-relevant hydrogen plasma loads. Measured by both electron backscatter diffraction over large fields of view and transmission Kikuchi diffraction of site-specific lamellae, such surface relief preferentially forms on grains with {110} planes parallel to the surface. This cannot be explained by the orientation-dependent resolved shear stress according to the Schmid law, threshold displacement energy anisotropy, or oxidation anisotropy. Furthermore, surface relief amorphization is revealed by high-resolution transmission electron microscopy imaging and selected area electron diffraction analysis, and is explained by a novel vacancy-induced amorphization mechanism. The results provide new insights into the thermal fatigue behavior of tungsten for fusion applications.

New method for reshaping of the WEST Lower Hybrid launcher to reach long pulse operation

The aim of WEST (tungsten-W Environment in Steady-state Tokamak) experiments is to master long pulses (1000s) and submit the ITER-grade tungsten divertor to power fluxes of 10 MW/m2. The lower hybrid current drive (LHCD) system, composed of a fully active multijunction (FAM-LH1) launcher and a passive active multijunction (PAM-LH2) launcher, with a total capability of 7 MW/1000s, is essential to meet these goals. The LH1 launcher was reshaped before the start of WEST to match the new plasma shape and thus improve its coupling capability. During the 2019–2021 experimental campaign, long pulse operation (∼ 55 s) was successfully and routinely performed with PLH = 3 MW and good LH coupling properties. Thanks to infrared monitoring during long pulses, the new toroidal shape of the LH1 launcher was found to avoid overheating on the front face edges. It also highlighted the existence of hot spots on the LH2 launcher's front face edges, limiting its power capability in high power and long pulse scenarios. A reshaping inside the vacuum vessel of the LH2 launcher was therefore performed during the shutdown in 2021. This paper describes the reshaping methods used for the two launchers and the tools developed for the reshaping of LH2 inside the tokamak.

Bubble Formation in ITER-Grade Tungsten after Exposure to Stationary D/He Plasma and ELM-like Thermal Shocks

Plasma-facing materials (PFMs) in the ITER divertor will be exposed to severe conditions, including exposure to transient heat loads from edge-localized modes (ELMs) and to plasma particles and neutrons. Tungsten is the material chosen as PFM for the ITER divertor. In previous tests, bubble formation in ITER-grade tungsten was detected when exposed to fusion relevant conditions. For this study, ITER-grade tungsten was exposed to simultaneous ELM-like transient heat loads and D/He (6%) plasma in the linear plasma device PSI-2. Bubble formation was then investigated via SEM micrographs and FIB cuts. It was found that for exposure to 100.000 laser pulses of 0.6 GWm−2 absorbed power density (Pabs), only small bubbles in the nanometer range were formed close to the surface. After increasing Pabs to 0.8 and 1.0 GWm−2, the size of the bubbles went up to about 1 µm in size and were deeper below the surface. Increasing the plasma fluence had an even larger effect, more than doubling bubble density and increasing bubble size to up to 2 µm in diameter. When using deuterium-only plasma, the samples showed no bubble formation and reduced cracking, showing such bubble formation is caused by exposure to helium plasma.

Defect evolution of neutron irradiated ITER grade tungsten after annealing

The microstructural evolution of neutron irradiated tungsten (W) after annealing and its correlation with the corresponding mechanical properties provides valuable insight on the defect interactions and their annihilation processes. This would result in the identification of recovery mechanisms, leading to the design of healing processes and thus, enabling the lifetime extension of fusion reactor components. Within this framework, samples from ITER specification forged W bar were neutron irradiated to a dose of 0.2 displacements per atom (dpa) at 600 °C in the Belgian Material Test Reactor (BR2) and subsequently annealed at 800 and 1000 °C. The evolution of the irradiation induced defects after annealing has been assessed by transmission electron microscopy (TEM), positron annihilation lifetime spectroscopy and electrical resistivity and its effect on the mechanical properties are discussed in terms of Vickers hardness. Neutron irradiation results in the formation of dislocation loops and voids. TEM observations show an increase in the size of both defect types after post irradiation (PI) annealing at both temperatures, accompanied by an initial increase in their density after PI annealing at 800 °C, followed by a subsequent decrease after PI annealing at 1000 °C. The presence of TEM-irresolvable defects of both types is revealed in the as-irradiated state, which is evidenced by the evolution of Vickers hardness, resistivity and positron lifetimes and their correlation after the PI annealing. In the as-irradiated state, as well as after PI annealing at both temperatures, the hardening is dominated by voids.

Fracture-mechanical behaviour of ITER grade tungsten subjected to three different rolling processes

We investigated the fracture-mechanical behaviour of three ITER grade tungsten plates subjected to different rolling processes. One plate was uniaxially rolled at a normal rolling ratio. The two other plates were subjected to cross rolling processes at normal and high rolling ratios. Two of the three plates were characterized for L-T and T-L orientations, whereas the plate subjected to the cross rolling at a normal rolling ratio was tested in the L-T orientation only. The DBTTs of the investigated plates determined via fracture-mechanical experiments were near 300°C, irrespective of the specimen machining directions. Marginally better fracture-mechanical behaviour was observed in the L-T orientation compared to T-L orientation for the uniaxially rolled plate. The plate subjected to the cross rolling process at the normal rolling ratio was found to have inferior fracture-mechanical behaviour. This manifested in the form of steep load drops in load-displacement curves, which persisted even at the highest test temperature of 800°C. Fractographic investigations revealed the governing failure modes of the investigated materials.

Damage behaviors in microstructures and mechanical properties of pure tungsten induced by repetitive thermal loads

Damage behaviors of ITER grade tungsten (W) induced by repetitive thermal loads have been investigated using an electron beam device EBMP-30. The metallographic observation, backscattered electron scanning electron microscopy, electron backscattered diffraction, atomic force microscopy analysis and tensile tests were carried out after the repetitive heat loads. Results indicate that there are negligible effects on the microstructures, surface morphologies and mechanical properties with the absorbed power densities (APD) lower than 10 MW/m 2 . While serious damages including the severe surface roughness with protruding structures along grain boundaries, and significant grains coarsening with the evolution of the grain shape occur undesirably due to the full recrystallization during the repetitive 30 MW/m 2 heat loads. Furthermore, after exposed to an APD of 30 MW/m 2 with 50 heat loads, the tensile properties deteriorate dramatically, resulting in a limited ultimate tensile strength of only 328 MPa and even a zero total elongation at 300 °C. The relationships between the thermal loads, microstructures evolution and degradation of mechanical properties have been analyzed.

Synergistic and separate effects of plasma and transient heat loads on the microstructure and physical properties of ITER-grade tungsten

Once ITER commences full power operation, the ITER divertor will be exposed to high thermal and particle loads. Tungsten was chosen as plasma facing material for the ITER divertor. It is, therefore, of the utmost importance to understand the behavior of ITER-grade tungsten under conditions similar to those it will have to withstand inside the reactor. In this study, ITER-grade tungsten samples were exposed to stationary D/He(6%) plasma and ELM-like transient heat loads in the linear plasma device PSI-2. The effects of each kind of load was first studied separately, and the synergistic effects obtained when exposed to both loads simultaneously were then analyzed. Additionally, the hardness of a recrystallized tungsten sample after exposure to simultaneous loads was tested via nanoindentation. The results indicate that hydrogen and helium embrittlement worsens the cracking behavior of the material when exposed to the simultaneous loads compared to only heat loads. Additionally, bubbles of up to 1 μm are formed under the surface due to the synergistic effects at the highest heat load. The nanoindentation tests showed that plasma and heat loads increase the hardness of the material by 39%, but only plasma loads appeared to have no effect on it.

The competing effects of temperature and neutron irradiation on the microstructure and mechanical properties of ITER grade tungsten

In the current work, ITER grade W irradiated at 1200 oC to a dose of 0.18 dpa has been investigated by Transmission Electron Microscopy (TEM), Positron Annihilation Spectroscopy (PAS), X-ray diffraction, impulse excitation and depth-sensing micro-indentation. The results are compared with those of annealed unirradiated W at 1200 oC. In the as fabricated material dislocations and small vacancy clusters of 1 to 3 vacancies are present. After annealing at 1200 oC, the small vacancy clusters disappear and only the dislocations with almost unchanged density remain. Also, hardness is reduced by 4% and creep increases by 13%. Neutron irradiation at 1200 °C results in the increase of the dislocation line density by 3.7 times and in the formation of dislocation loops and voids. The irradiation also causes a hardness increase of about 20% and a creep decrease of 15%. Both the elastic and shear moduli decrease by 4% after irradiation at 1200 °C contrary to the effect of only annealing at this temperature which causes an increase in their values by 4%. Texture changes are observed only after neutron irradiation.

Recrystallization-mediated crack initiation in tungsten under simultaneous high-flux hydrogen plasma loads and high-cycle transient heating

Tungsten and tungsten-based alloys are the leading material choices for the divertor plasma facing components (PFCs) in future fusion reactors. Recrystallization may occur when they undergo high heat loads, drastically modifying the predesigned grain structures and the associated desired mechanical properties. However, the influence of recrystallization on the thermal fatigue behavior of tungsten PFCs still remains unclear. In this study, ITER-grade tungsten was simultaneously exposed to a high-flux hydrogen plasma (∼5 × 1024 m−2 s−1) and high-cycle (104–105) transient heat loads in the linear plasma device Magnum-PSI. By correlating the surface temperature distribution, obtained by analyzing temperature-, wavelength-, and surface-dependent emissivity, and the surface modifications of the plasma exposed specimens, the crack initiation heat flux factor threshold was found to be ∼2 MW m−2 s0.5 (equivalently, ∼0.07 MJ m−2 for a 1 ms pulse). Based on electron backscatter diffraction analyses of cross-sections near the crack initiation sites, faster recrystallization kinetics near the surface compared to literature was observed and the surface cracks preferentially initiated at high angle grains boundaries (HAGBs). Upon recrystallization, the yield strength decreases which entails increasing cyclic plastic strains. The HAGBs fraction is increased, which constrains the transfer of plastic strains at grain boundaries. The recrystallization decreases the dislocation density, which promotes heterogeneous deformation. All these mechanisms explain the reduced crack initiation threshold of recrystallized tungsten compared to its as-received counterpart. The results provide new insights into the structural failure mechanisms in tungsten PFCs exposed to extreme fusion plasmas.

Deuterium transport in ITER-grade tungsten

The transport behavior of hydrogen isotopes in ITER grade tungsten (IG-W) and commercial rolled polycrystalline tungsten (CR-W) was investigated using deuterium gas. Analyses using field-emission scanning electron microscopy revealed that both IG-W and CR-W had a polycrystalline structure, but their surface grain structures were apparently different. The permeability and diffusivity of deuterium in IG-W and CR-W were obtained using a time-dependent gas-phase technique in the temperature range of 650–850 °C, being observed to be affected by surface morphology and grain boundaries. Our results in this study were also compared with previous results for polycrystalline W reported by other authors.

High-heat flux tests of tungsten divertor mock-ups with steady-state plasma and e-beam

Abstract Mock-ups of the ITER divertor plate are tested with the combination of steady-state plasma and e-beam loads: (1) thermocyclic tests with powerful electron beam load of up to 49 MW/m2 in the e-beam facility; and then (2) subsequent testing in the PLM plasma device with steady-state plasma loads of 0.5–1 MW/m2 and more. Such tests simulate the variable load on divertor plates in the ITER. Tungsten samples are irradiated with helium plasma in experiments on the PLM plasma device with discharge duration of up to 200 min. The tests of ITER-grade tungsten VM-P with the combination of steady-state plasma and thermocyclic e-beam loads led to erosion, cracking, and nanostructured “fuzz” structure growth on the material surface. Post-mortem scanning electron microscopy revealed a stochastic nanostructured surface with dimension of structural elements less than 100 nm. The growth of a nanostructured surface with a “fuzz”-type structure and high porosity is observed. The results of such tests are of interest to estimate the erosion of tungsten in fusion reactors, including ITER, fusion neutron source FNS and DEMO.

High-heat flux tests of fusion materials with stationary plasma in the PLM device

Abstract The PLM plasma device was constructed for high heat flux tests of fusion plasma-facing materials and in-vessel components of a fusion reactor. The ITER-grade tungsten samples were irradiated with steady-state plasma in PLM. The combined tests of ITER-grade tungsten samples with an e-beam load of 40 MW/m2 and stationary steady-state plasma load of ∼1 MW/m2 led to erosion, cracking, and nanostructured “fuzz” structure growth on the material surface. The capillary porous system of liquid tin was tested with steady-state plasma in PLM during ∼200 min demonstrating sustainability under high heat plasma load. Lithium materials deposited in the T-10 tokamak during experiments with lithium capillary-porous system have been irradiated with stationary steady-state plasma in PLM to test the evolution of the deposits under long-term plasma load.

Mechanical properties of three kinds of ITER-Grade pure tungsten with different manufacturing processes

• The Vickers hardness test results showed no significant data dispersion. • Micro pillar test results were some data dispersion identified in both the center and the near-surface. • The tensile test showed comparably large data dispersion. • Weak defect structure may complicate the cracks, resulting in the data dispersion. Pure tungsten plates with three kinds of manufacturing processes, i.e., ITER-Grade tungsten, cross-rolled tungsten with a moderate rolling ratio and another cross-rolled tungsten with a high rolling ratio, were evaluated by the Vickers hardness test, tensile test at room temperature and 500 °C, and micro pillar test. The Vickers hardness test results showed that the hardness value of all types of the tungsten plates was 410–430 Hv and no significant data dispersion was identified. However, the tensile test showed comparably large data dispersion, which was particularly drastic at room temperature. In parallel, to determine the effect of the specimen cut location thorough the plate thickness direction (S-Direction), micro pillar tests were conducted. Although there were some data dispersion, test results showed plastic deformation and no marked effect of the cut location was identified. It is therefore believed that the specimen cut location did not affect the data dispersion identified in the tensile test. Alternatively, the probable explanation for such large dispersion is partial detachment of the surface bulk structure. Such specific microstructure can make crack paths more complicated, resulting in the data dispersion.

Deuterium retention in tungsten and reduced activation steels after 3 MeV proton irradiation

Abstract Nuclear fusion plasma-facing materials (PFM) will suffer from irradiation, leading to significant changes in the material properties. This study investigates the impact of displacement damage on the deuterium retention near room temperature. ITER grade tungsten, Eurofer-97, and HiperFer 17Cr5 steel samples are irradiated with a tandem accelerator with ~3 MeV protons at currents of 100–600 nA on 250–550 µm spots at 320±10 K. In total 33 spots from 0 to 0.9 displacements per atom (DPA) at 0–4 µm depth are irradiated on 5 samples. After irradiation, the samples are exposed to D2 plasmas with a peak ion-flux of 2.1 × 1021 D/m²s for 4 h at The long-term D retention in both W and steel increases with DPA with a saturation starting around 0.2 DPA. Retention in W increased by a factor 12 with up to 3.2 at.% D, while in steel increases up to 180 times with up to 0.08 at.% D were observed. The results highlight the importance of using steels also in PFMs. Compatibility of the results with heavy ion irradiations boosts the confidence in inter-comparability between different ion types, but also between ions and neutrons.

Synergistic effects of particle and transient heat loads on ITER-grade tungsten

ITER’s divertor will have to sustain high particle and heat loads. Tungsten has been chosen as plasma facing material for the entirety of ITER’s divertor. Therefore, it is important to study the behavior of ITER-grade tungsten under fusion relevant loads. In this work, tungsten samples with two different microstructures were studied, one with needle-like grains transversal to the surface, and the other with larger, isotropic grains obtained after recrystallization. The samples were exposed to a steady state plasma and a pulsed laser in the linear plasma device PSI-2 to simulate the conditions of the divertor in a fusion reactor. The results show that the transversal samples have a higher damage threshold, but at higher power density and pulse numbers, the damage observed, as well as the surface roughness, is similar. The recrystallized samples displayed a shallower crater depth, which might be an important factor to be considered when selecting a material for ITER and should be studied further.

Reduced D trapping by plasma-implanted He nanobubbles in radiation damaged tungsten

The presence of a plasma-implanted helium nanobubble layer significantly reduces deuterium (D) retention in undamaged commercial ITER grade tungsten (W). In this paper, we show evidence that this phenomenon can survive displacement damage. A He plasma exposure (sample temperature 643 K, ion flux 1022 m−2 s−1 at 100 eV, fluence 1025 m−2) pre-treatment was performed to create a thin He nanobubble layer in the first ~15 nm of ITER grade W samples. Samples were then irradiated by 5 MeV Cu ions at room temperature to create 0.001 to 0.1 dpa with peak damage rates occurring about 860 nm below the sample surface. Samples without He plasma exposure pre-treatment were also irradiated by 5 MeV Cu ions to provide a controlled baseline. All samples were subsequently exposed to D plasma at 373 K to a fluence of 1024 m−2. Nuclear reaction analysis results show that across a range of peak dpa ranging from 0.001 to 0.1 dpa, D retention inventory in the samples with He plasma exposure pre-treatment is reduced by a factor of two compared to samples without He plasma exposure pre-treatment. Transmission electron microscopy directly showed a surviving nanobubble layer after 0.1 dpa damage. However, the thickness of the bubble layer appears to have been reduced. The results suggest that the plasma-implanted He nanobubble layer can survive radiation damage and still function to reduce D diffusion and retention in tungsten-based plasma facing components.

Improved tritium retention modeling with reaction-diffusion code TMAP and bulk depth profiling capability

Abstract ITER-grade tungsten (W) specimens were exposed to similar deuterium (D) plasma condition (ion flux density of 6.0 × 1021 D m − 2s−1, D ion fluence of 5.0 × 1025 D m − 2) at the surface temperature of 623 K. Thermal desorption spectroscopy was used to measure total D retention in one W specimen after D plasma exposure. Glow-discharge optical emission spectroscopy was used to measure D depth profiling from the other W specimen exposed to the similar condition, and deep D trapping up to 25 μm was observed. When the normalized D depth profile was used with a reaction-diffusion code TMAP7 to model experimental D desorption behavior, an excellent agreement to experimental results was achieved. The modeling results suggested that the predominate mechanism of the deep D trapping might be D trapping in intrinsic intergranular cracks in ITER-grade W.

Thermal desorption spectroscopy of high fluence irradiated ultrafine and nanocrystalline tungsten: helium trapping and desorption correlated with morphology

Microstructural changes due to displacement damage and helium desorption are two phenomena that occur in tungsten plasma facing materials in fusion reactors. Nanocrystalline metals are being investigated as radiation tolerant materials that can mitigate these microstructural changes and better trap helium along their grain boundaries. Here, we investigate the performance of three tungsten grades (nanocrystalline, ultrafine and ITER grade tungsten), exposed to a high fluence of 4 keV helium at both RT and 773 K, during a thermal desorption spectroscopy (TDS) experiment. An investigation of the microstructure in pre-and post-TDS sample sets was performed. The amount of desorbed helium was shown to be highest in the ITER grade tungsten and lowest in the nanocrystalline tungsten. Correlating the desorption spectra and the microstructure (grain boundaries decorated with nanopores and crack formation) and comparing with previous literature on coarse grained tungsten samples at similar irradiation and TDS conditions, revealed the importance of grain boundaries in trapping helium and limiting helium desorption up to a high temperature of 1350 K in agreement with transmission electron microscopy studies on helium irradiated tungsten which showed preferential and large facetted bubble formation along the grain boundaries in the nanocrystalline tungsten grade.

Review of recent experiments on the T-10 tokamak with all metal wall

Transition to an all-metal wall was realized on T-10 by replacement of graphite limiters with tungsten ones. Light impurity levels remained high and W accumulation in the plasma core was revealed. A movable lithium (Li) limiter was added to investigate the possibility of the limitation of tungsten and light impurity levels in plasma. For the first time, tokamak results on tungsten protection with Li were obtained in OH and ECRH regimes. After lithization the tungsten density in the core dropped more than an order of magnitude, while W influx into plasma decreased 2–4 times. Drastic drops of light impurities in plasma were observed together with improvement of energy confinement time and density limit values. Nevertheless, Li levels in plasma remained low in both OH and ECRH regimes. Li density in the core as low as 0.5% of ne was obtained. Tungsten transport in T-10 plasma was investigated and results on prevention of W accumulation with central ECRH were obtained. Effects of plasma exposure on ITER-grade tungsten plates from limiters were studied. Investigations of density fluctuation with correlation reflectometry confirmed a decrease of fluctuation amplitude on high field side. Modeling showed that this effect can be, to a great extent, explained by the non-locality of reflectometry. Toroidal correlations at a distance of 2.5 m along field lines were studied. Three-wave interaction between geodesic acoustic modes and broad-band turbulence was found by analysis of heavy ion beam probe diagnostics data. The possibility of plasma current control and the prevention of non-thermal electron beams formation at density limit disruption by means of ECR heating and the controlled operation of OH power supply system has been demonstrated. The study of plasma density decay after gas puff switch off during density ramp-up phase in OH regimes and the effect of ‘density pump out’ during ECRH showed that both effects can be explained by the assumption regarding electron transport growth upon reaching marginal pressure profile. A new type of internal transport barrier in the narrow zone near q = 1 surface has been found.