Behavior of tungsten under irradiation and plasma interaction

Rieth, M.; Doerner, R.; Hasegawa, Akira; Ueda, Y.; Wirtz, M.

doi:10.1016/j.jnucmat.2019.03.035

Cited by 151 publications

(60 citation statements)

References 568 publications

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“…The CVD-W coatings with a high purity larger than 99.9999% have a typical columnar crystal structure. The density of CVD-W is 99% that of bulk W, and it is 19.23 g/cm 3 . The purity and density values of CVD-W are supplied by the manufactures (in Xiamen).…”

Section: Sample Preparationmentioning

confidence: 99%

“…However, there are still a series of problems during the process of commercialization and engineering application of nuclear fusion energy, a key one of which is to choose the appropriate plasma-facing materials (PFMs) [1]. So far, W (tungsten) has been considered as the most promising candidate material for the first wall and divertor material in the international experimental fusion reactor (ITER) and demonstration fusion reactor (DEMO) due to its high melting point, high thermal conductivity, low sputtering rate for light elements, and low activation after neutron irradiation [2,3]. In the operation of fusion reactor, the PFMs are subjected to extremely severe conditions: the high flux plasma particle bombardment, high energy neutron irradiation, steady-state thermal shock, and transient thermal shock [1,4].…”

Section: Introductionmentioning

confidence: 99%

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Evolution of Defects in CVD-W Irradiated by H/He Neutral Beam Using Positron Annihilation Spectroscopy

Xia

Liu

Gong

et al. 2021

Metals

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One of the key problems for the application of nuclear fusion energy is to select the suitable plasma facing materials (PFMs). Among the W-based materials, CVD-W exhibits some unique advantages. In order to estimate the performance of CVD-W under the fusion environment, the vacancy-type defects and their evolution are investigated by the Doppler-broadening slow positron beam analysis (DB-SPBA) combined with SEM (scanning electron microscope). There are two kinds of neutral beam irradiation, the pure H neutral beam and the H + 6 at.% He neutral beam irradiation, which are performed at the neutral beam facility GLADIS (IPP, Germany). The surface temperatures of CVD-W irradiated by H (H + 6 at.% He) are 850 and 1000 (700 and 800 °C). By comparing the samples under different conditions, the defect evolution of CVD-W is obtained. As for the pure H neutral beam irradiated samples, the DB-SPBA results demonstrate that the CVD-W sample at the surface temperature of 1000 °C, compared to the 850 °C sample, shows a decrease in S parameters, which is due to the reduction of vacancy-type defect concentration. The defect damage layer in 1000 °C sample is narrower than that of 850 °C sample and the defect type tends to be consistent in 1000 °C sample. The SEM results suggest that the surface damage of the 1000 °C sample was recovered to some extent. As for the H + 6 at.% He neutral beam irradiated samples, compared with the CVD-W sample at the surface temperature of 700 °C, the 800 °C sample shows an increased S parameters, which can be attributed to the volume increase of vacancy-type defect. The defect damage layer in the 800 °C sample is wider than that of the 700 °C sample. Both the H + 6 at.% He irradiated samples show complex defect types. The surface of the 800 °C sample exhibits more dense pinhole damage structures compared to that of the 700 °C sample.

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Section: Sample Preparationmentioning

confidence: 99%

Section: Introductionmentioning

confidence: 99%

Evolution of Defects in CVD-W Irradiated by H/He Neutral Beam Using Positron Annihilation Spectroscopy

Xia

Liu

Gong

et al. 2021

Metals

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“…Plasma-facing components (PFCs) for future fusion reactors will have to withstand extremely harsh conditions involving high heat fluxes (both steady-state and thermal shocks) and bombardment of plasma species (ions, electrons, neutral atoms, and high-energy neutrons) [ 1 ]. Tungsten is considered as the prime candidate material for these components, particularly for its refractory nature (high melting point and high strength at elevated temperatures), high resistance to sputtering, good thermal conductivity, etc., [ 2 , 3 , 4 ]. However, joining of tungsten to steel- or copper-based structural or cooling system presents a significant challenge.…”

Section: Introductionmentioning

confidence: 99%

W + Cu and W + Ni Composites and FGMs Prepared by Plasma Transferred Arc Cladding

2021

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Tungsten-based materials are the most prospective candidates for plasma-facing components of future fusion devices, such as DEMO. W-based composites and graded layers can serve as stress-relieving interlayers for the joints between plasma-facing armor and the cooling or structural parts. Coating/cladding techniques offer the advantages of eliminating the joining step and the ability to coat large areas, even on nonplanar shapes. In this work, W + Cu and W + Ni composites were prepared by pulsed plasma transferred arc (PTA) cladding on several different substrates. Optimization of the process was carried out with respect to powder mixture composition and process parameters like arc current, plasma gas composition, and traverse velocity. Dense claddings of several millimeters thickness and various W content were achieved. Moreover, multilayers with W content gradually varying from 47 to 92% were formed. The structure, compositional profiles, and thermal properties of the claddings were characterized.

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“…Besides renewable power generation methods, such as for example concentrated solar power, future fusion reactors will greatly benefit in particular by the use of W in plasma-facing structural parts. In the latter case, the material has to withstand very high operation temperatures as well as it has to tolerate a significant amount of neutron-induced radiation damage [1][2][3] . A detailed experimental understanding of the neutron damage processes on the nanoscale, which is still limited at present time, is of utmost importance in the material qualification process.…”

Section: Introductionmentioning

confidence: 99%

“…Basically, there are three mechanisms that may degrade the beneficial material properties of W during neutron irradiation: (1) Firstly, this includes the formation of lattice defects such as Frenkel pairs, interstitial and vacancy clusters and dislocation loops. (2) In addition, the formation of voids leads to swelling and radiation hardening that would restrict the lifetime of components beyond certain design limits. (3) The transmutation processes also lead to the formation of rhenium-(Re) or osmium-(Os) rich phases, while helium transmutation could stabilize voids.…”

Section: Introductionmentioning

confidence: 99%

New microstructural insights into neutron-irradiated tungsten

Dürrschnabel¹,

Klimenkov²,

Jäntsch³

et al. 2020

Preprint

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The development of appropriate materials for fusion reactors that can sustain high neutron fluencies at elevated temperatures remains a great challenge. Tungsten is one of the promising candidate materials for plasma-facing components of future fusion reactors, due to several favorable properties as for example a high melting point, a high sputtering resistivity, and a low coefficient of thermal expansion. The microstructural details of a tungsten sample with a 1.25 dpa (displacements per atom) damage dose after irradiation at 800°C were examined by transmission electron microscopy (TEM). Three types of radiation-induced defects were observed, analyzed and characterized: (i) voids with sizes ranging from 10 nm to 65 nm, (ii) dislocation loops with a size of up to 10 nm and (iii) W-Re-Os containing σ- and χ-type precipitates. The distribution of voids as well as the nature of the occurring dislocation loops were studied in detail. In addition, nano-chemical analyses revealed that the σ- and χ-type precipitates, which are sometimes attached to voids, are surrounded by a solid solution cloud enriched with Re. For the first time the crystallographic orientation relationship of the σ- and χ-phases to the W-matrix was specified. Furthermore, electron energy-loss spectroscopy could not unambiguously verify the presence of He within individual cavities.

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Behavior of tungsten under irradiation and plasma interaction

Cited by 151 publications

References 568 publications

Evolution of Defects in CVD-W Irradiated by H/He Neutral Beam Using Positron Annihilation Spectroscopy

Evolution of Defects in CVD-W Irradiated by H/He Neutral Beam Using Positron Annihilation Spectroscopy

W + Cu and W + Ni Composites and FGMs Prepared by Plasma Transferred Arc Cladding

New microstructural insights into neutron-irradiated tungsten

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