Deuterium behavior in tungsten exposed to deuterium plasma with rising or declining temperature

Wang, Ting; Yuan, Yue; Guo, Wangguo; Ma, Xiaolei; Liu, Mi; Wang, Jun; Cheng, Long; Zhu, Xiu-Li; Lü, Guang-Hong

doi:10.1016/j.jnucmat.2020.152243

Cited by 13 publications

(3 citation statements)

References 54 publications

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“…The measurement area for surface blister statistics was 4.62 × 10 5 µm 2 for each sample, and specific details can be found in table 2. The statistics on the surface blister encompass intergranular blisters and intragranular blisters which are the typical blistering types of recrystallized W [27,47].…”

Section: Surface Morphologies After Deuterium Plasma Exposurementioning

confidence: 99%

Deuterium retention in cyclic transient heat loaded tungsten with increasing cycle numbers

Ren,

Yuan,

Feng

et al. 2024

Nucl. Fusion

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Surface damage and microscopic defect evolution of tungsten (W) armor under transient heat loads are key factors for fuel retention in fusion reactors. In this work, experiments were conducted to investigate the effects of cyclic thermal shocks on deuterium (D) retention and surface blistering in W. Thermal shock experiments were conducted on recrystallized W using an electron beam with a power density of 0.15 GW m-2 across 100-1500 cycles, followed by D plasma exposure with high-fluence (~ 1 × 1026 D m-2). The results demonstrate that samples subjected to 500 and 1500 cycles exhibit a significant presence of sub-grains within 90 μm. Notably, the inhibition of blistering induced by thermal shock leads to a substantial reduction in D retention (5.45 × 1019 D m-2) at lower cycle numbers (100 cycles) compared to the reference sample (2.35 × 1020 D m-2) which was only exposed to D plasma. When cycle numbers increase to 500 and 1500, D retention reaches 1.98 × 1020 D m-2 and 4.56 × 1020 D m-2, respectively. Based on the Tritium Migration Analysis Program, we propose that total D retention is a consequence of the competition between defects reduced by thermal shock-induced suppression of blistering and defects generated by plastic deformation induced by thermal stress. D retention initially decreases with the increase in cycle numbers, followed by a subsequent rise, with the inflection point slightly higher than 500 cycles. Additionally, due to the extensive scope of thermal stress, an escalated exposure period will result in substantial D captured by heat-induced defects, consequently intensifying the D retention. Whether there exists an upper limit to D retention induced by the increasing thermal shock cycles necessitates further experimental analysis. Nonetheless, it is evident that thermal shock significantly contributes to D retention within a profoundly deep bulk region under high cycles.

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Section: Surface Morphologies After Deuterium Plasma Exposurementioning

confidence: 99%

Deuterium retention in cyclic transient heat loaded tungsten with increasing cycle numbers

Ren,

Yuan,

Feng

et al. 2024

Nucl. Fusion

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“…Tungsten's significance in studying plasma-material interactions has been pivotal for recognizing potential operational issues within ITER [3]. Temperature [4], neutron irradiation [5], and hydrogen (H) or helium (He) ion implantation [6][7][8] are crucial factors influencing embrittlement in W. Understanding not only the influence of sample temperature and plasma composition on nanobubble formation but also identifying potential synergies among these factors is critical for predicting the performance of tungsten components in the ITER divertor. Numerous investigations have been conducted over the past two decades to understand the behaviour of H and He isotopes in W, especially on their roles in inducing bubble growth during exposure to H/He plasma with energy levels of 20 eV-200 eV [9][10][11].…”

Section: Introductionmentioning

confidence: 99%

Temperature-dependent bubble growth under synergistic interactions of hydrogen and helium in tungsten

Niu,

Qin,

Suman

et al. 2024

Nucl. Fusion

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A novel theoretical model based on modified diffusion rate equations is proposed to simulate the retention of hydrogen isotopes and the dynamics of bubble growth in tungsten (W) when exposed to simultaneous hydrogen (H) and helium (He) plasma irradiations. Simulation is conducted to assess the influence of temperature as well as simultaneous H and He irradiation at an increasing fluence. Not only to develop a holistic understanding but also to substantiate simulation findings about synergy between H and He plasma irradiation, a W sample is exposed sequentially to H and He plasma at 873 K using the large-power material irradiation experimental system (LP-MIES). The topographical changes in the W sample are investigated using atomic force microscopy (AFM) after each plasma irradiation exposure sequence. Simulation results reveal that the ability of a bubble containing both H and He to trap adjacent H/He atoms is primarily governed by their individual partial pressure within the bubble. Furthermore, at elevated temperatures, the synergy between H and He significantly enhances the retention of H isotopes in W. AFM micrographs of the W sample exposed to both H and He plasma irradiation show a severely damaged and locally delaminated layer, absent in the sample exposed only to either H or He, conclusively establishing evidence of synergy between H and He irradiation effects. The average bubble radius computed using the model aligns excellently with experimentally determined values obtained through SEM/AFM analysis. The robustness of the proposed model is also assessed by comparing bubble radius and H isotopes retention at various temperatures with experimental data reported in the literature.

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“…When the incident energy is much lower than the displacement damage threshold, the resulting supersaturated hydrogen in the material may lead to serious blistering and surface modification [24,25]. According to the interaction between tungsten and hydrogen plasma [26][27][28], the irradiation damage on the surface of the OFC material might be also strongly dependent on the material temperature, hydrogen ion fluences and hydrogen ion energy. Moreover, copper was irradiated with low energy hydrogen ions from 30 • C to 300 • C, and the microstructure evolution of copper was observed under an in situ transmission electron microscope in earlier study [29].…”

Section: Introductionmentioning

confidence: 99%

Stress-driven surface swell and exfoliation of copper as the plasma-facing materials in NBI ICP source

Cui¹,

Fan²,

Niu³

et al. 2021

Plasma Phys. Control. Fusion

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Neutral beam injection (NBI) heating is a significant auxiliary heating method used in Tokamak fusion devices. The material of faraday shield (FS) and accelerator grids in the NBI inductively coupled plasma (ICP) source can be damaged during operation by the high-density hydrogen plasma irradiation, and thus affecting the stability of the NBI system. In this paper, a series of hydrogen plasma exposure experiments are performed on oxygen-free copper (OFC) specimens at 400 K–850 K with ion energy of 20–200 eV and irradiation fluence up to 1.0 × 1025 m−2. Meanwhile, the rate equation model is adopted for numerical simulation of the bubble growth and hydrogen retention. The influence of OFC surface temperature, hydrogen ion energy and fluence on OFC damage are experimentally and numerically investigated. Surface observations show that swell and exfoliation are formed on the OFC samples at 400 K and 600 K by scanning electron microscopy. The hydrogen ion energy varying from 20 to 200 eV at 400 K is observed to have little effect on OFC surface microstructure. The simulative results show that there exist different critical temperatures when the initial bubble radius changes. The bubble surface density rises and the bubble size decreases with increasing temperature (below the critical temperature). In addition, adjacent bubbles get closer to each other with the growth of hydrogen bubbles, and the strong tensile stress is produced inside the surrounding material of hydrogen bubbles. Some cracks caused by hydrogen bubbles appear on the surface of the OFC to relax the pressure-induced stress, ultimately leading to OFC FS/grids material damage. This investigation helps to understand hydrogen retention and failure mechanisms of OFC materials under extreme operation conditions in the NBI devices.

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Deuterium behavior in tungsten exposed to deuterium plasma with rising or declining temperature

Cited by 13 publications

References 54 publications

Deuterium retention in cyclic transient heat loaded tungsten with increasing cycle numbers

Deuterium retention in cyclic transient heat loaded tungsten with increasing cycle numbers

Temperature-dependent bubble growth under synergistic interactions of hydrogen and helium in tungsten

Stress-driven surface swell and exfoliation of copper as the plasma-facing materials in NBI ICP source

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