New insights into microstructure of neutron-irradiated tungsten

Dürrschnabel, Michael; Klimenkov, M.; Jäntsch, U.; Rieth, M.; Schneider, Holger; Terentyev, D.

doi:10.1038/s41598-021-86746-6

Cited by 51 publications

(17 citation statements)

References 60 publications

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“…A detailed understanding of the interaction of W with the fusion products (helium and neutron) and the fusion fuel (deuterium (D) and tritium) is needed, especially because tritium is a scarce and radioactive element. Tungsten exposure to fusion products generate morphology changes in the bulk of W, as numerous studies have shown [1][2][3][4], and the influence of such bulk defects onto fusion fuel retention is presently an active field of research [5,6]. In contrast, pure surface effects are often neglected in plasmawall interaction studies and global recycling coefficients are used for simulating edge plasma physics [7].…”

Section: Introductionmentioning

confidence: 99%

Surface oxygen versus native oxide on tungsten: contrasting effects on deuterium retention and release

et al. 2022

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We performed a direct comparison of deuterium retention and release from tungsten in presence or in absence of oxygen impurities. A single crystal of W(110) was used to prepare tungsten with four different surface states: with its native oxide, atomically clean, covered with half a monolayer of oxygen atoms, and covered with three fourths of a monolayer of oxygen atoms. For a D ion fluence of 3×1021 D+ m-2 implanted at 300 K, deuterium retention was highest with the native oxide, lowest with three fourths of a monolayer of oxygen atoms at the surface and intermediate for the clean surface. This counterintuitive result is explained by a different localization of deuterium retention in these samples. For tungsten with its native oxide, deuterium retention occurs solely in the bulk, i.e. below the first atomic plane of the surface. For clean tungsten, deuterium retention occurs in part at the surface and sputtering should play a role. For tungsten with a sub-monolayer surface coverage of oxygen atoms, a transition from surface to bulk retention is observed above half a monolayer of adsorbed oxygen. Striking differences in desorption peak(s) temperature(s) are observed between D ion-implanted samples and D2 molecules-exposed samples. These results highlight the importance of the (near-) surface localization of oxygen and deuterium on the temperature dependence of deuterium desorption rate i.e. on the fusion fuel recycling coefficient.

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Section: Introductionmentioning

confidence: 99%

Surface oxygen versus native oxide on tungsten: contrasting effects on deuterium retention and release

et al. 2022

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“…Secondly, He peak may appear shifted towards the W plasmon in the overpressurized bubbles [39,61,58]. Lastly, a presence of impurities such as oxygen in the substrate (e.g., oxidation) or inside the bubble may add another contributions to the spectrum [53].…”

Section: Density and Gas Pressure In Helium Bubblesmentioning

confidence: 99%

Helium-induced morphology evolution in tungsten under thermal treatment

Ialovega¹,

Bernard²,

Barthe³

et al. 2022

Nucl. Fusion

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Surface and near-surface morphology evolution of helium-irradiated tungsten due to thermal cycling up to the ITER-relevant temperature of 1350 K was studied using electron microscopy and positron annihilation spectroscopy techniques. Recrystallized polycrystalline tungsten samples were irradiated with 75 eV helium plasma at the surface temperature of 1073 K and the fluence of 3×1023 He/m-2 which induced formation of holes on the surface and bubbles in the near-surface of tungsten. Subsequent annealing experiments were combined with a detailed electron microscopy analysis to investigate the shape and density changes of holes and helium bubbles with respect to grain orientation. We show that the initially circular holes and round bubbles became faceted upon heating with 1 K/s ramp up to 870 K. Low energy - low fluence deuterium implantation did not play a role in bubble faceting. Annealing cycles up to 1350 K induced 5 nm subsurface annealing resulting in bubbles removal and surface smoothing. Electron energy loss spectroscopy measurements allowed estimation of helium gas density and pressure inside bubbles. Positron annihilation spectroscopy allowed to investigate the nature of defects and their evolution in the helium-irradiated tungsten with thermal cycling.

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“…Over the past decades, extensive microstructural characterizations of W alloys have already been carried out to understand the microstructural response to neutron irradiation [7,[9][10][11][12]. It can be concluded that neutron irradiation can cause the formation of lattice defects such as dislocation loops, Frenkel pairs and vacancies in the W alloys [13,14], and Materials 2022, 15,1985 2 of 14 can also lead to the formation of rhenium-(Re)-or osmium-(Os)-rich precipitates due to the trigger transmutation process [15,16]. However, owing to low irradiation intensity, long experimental periods, radioactive problems and resource shortages, researchers mostly used different kinds of heavy ion irradiation to screen tungsten materials because of the more obvious damage effect as well as the flexible control for experiment parameters [17,18].…”

Section: Introductionmentioning

confidence: 99%

Superior Radiation Resistance of ZrO2-Modified W Composites

et al. 2022

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The microstructure and mechanical properties of pure W, sintered and swaged W-1.5ZrO2 composites after 1.5 × 1015 Au+/cm2 radiation at room temperature were characterized to investigate the impact of the ZrO2 phase on the irradiation resistance mechanism of tungsten materials. It can be concluded that the ZrO2 phase near the surface consists of two irradiation damage layers, including an amorphous layer and polycrystallization regions after radiation. With the addition of the ZrO2 phase, the total density and average size of dislocation loops, obviously, decrease, attributed to the reason that many more glissile 1/2<111> loops migrate to annihilate preferentially at precipitate interfaces with a higher sink strength of 7.8 × 1014 m−2. The swaged W-1.5ZrO2 alloys have a high enough density of precipitate interfaces and grain boundaries to absorb large numbers of irradiated dislocations. This leads to the smallest irradiation hardening change in hardness of 4.52 Gpa, which is far superior to pure W materials. This work has a collection of experiments and conclusions that are of crucial importance to the materials and nuclear communities.

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New insights into microstructure of neutron-irradiated tungsten

Cited by 51 publications

References 60 publications

Surface oxygen versus native oxide on tungsten: contrasting effects on deuterium retention and release

Surface oxygen versus native oxide on tungsten: contrasting effects on deuterium retention and release

Helium-induced morphology evolution in tungsten under thermal treatment

Superior Radiation Resistance of ZrO2-Modified W Composites

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