Irradiation hardening induced by blistering in tungsten due to low-energy high flux hydrogen plasma exposure

Chen, W.Q.; Xiao, Xiazi; Pang, Bo; Si, S.; Jia, Y.Z.; Xu, Baosheng; Morgan, T.W.; Liu, W.; Chiu, Yu-Lung

doi:10.1016/j.jnucmat.2019.05.004

“…In physical chemistry sense, He atoms have strong repulsion to W atoms [80,92]. This ultra-low solubility forces He atoms to self-precipitate into small He bubbles [83] that become nucleation sites [90] for further void growth [93] under radiation induced vacancy supersaturations [94], resulting in material swelling [69,86,95] and high temperature He embrittlement [71,96,97], as well as surface blistering [75][76][77][78] under low energy and high flux He bombardment [54,98] at elevated temperatures [99]. This may be mitigated by engineering structures in material which help in outgassing of He.…”

Section: Bubblementioning

confidence: 99%

“…It has high melting point, low activation, good thermo-mechanical properties, good weldability, good corrosion resistance, low sputter erosion/redeposition, low tritium retention/co-deposition and aforementioned nuclear properties. However, upon exposure to plasma, it shows rapid surface degradation exhibiting surface blisters [75][76][77][78] and formation of under dense nanostructure -fuzz as described above. It is supposed to happen as W has bcc structure with 12 small tetrahedral interstitial sites and 6 large octahedral interstitial sites [100,174] with lower atomic packing factor and coordination number than fcc (bcc 8, fcc 12) [175], small cation size and a higher tendency for external interstitial atom to occupy 12 small tetrahedral sites [175,176], poor tendency for host / self (W) atom to occupy 6 large octahedral sites, thus poor tendency for self-interstitial solid solution to form and strengthening to occur.…”

Section: Self-interstitial Solid Solution Strengthening -Crystallogramentioning

confidence: 99%

“…To this end, various materials (such as W [47][48][49][50], addition of Rh in W [51], use of bcc Fe [52,53], Ta [54], W-Ta [55], Ta/Fe [56], Pd [57], nanocrystalline Cu [58], SiOC/Crystalline Fe nanocomposite [59], W-K [60], reduced activation steel [61], ferritic [62], ferritic/martensitic steels [63], Be pebbles [64][65][66][67], Be and beryllides [68], graphite, carbon fiber composite [69]) and high Z atoms (Zr, No, Mo, Hf, Ta) [70] have been tested but none proved satisfactory [71][72][73][74]. All show rapid surface degradation exhibiting surface blisters [75][76][77][78] and formation of fuzz [51,[79][80][81][82] or under dense nanostructure [40] after bubble. These rapidly degrade their physical (e.g.…”

Section: Introductionmentioning

confidence: 99%

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Plasma Facing Material by Self-Interstitial Solid Solution Strengthening – Problem, Proposition, and a Solution

Rafique¹

2020

Preprint

0

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Bubble (point defect) – a precursor of fuzz or under dense nanostructure formation is crystal lattice defect. Suitable selection of crystal lattice which inhibit Frenkel pair generation and intrinsically promotes selfinterstitial solid solution strengthening contributes effectively towards making plasma facing material. For this, interstitial sites, their size, amount / fraction, positions, tendency of occupation and diffusion parameters (e.g. activation energies (Q), activation volumes) are determined. Fcc iron carbon alloys (austenitic stainless steels AISI / SAE 321, fcc structure, Pearson code cF4, space group Fm3̅m) are proposed as suitable candidates. Along with their room temperature fcc structure having 12 interstitial positions (4 octahedral, 6 coordination sites and 8 tetrahedral, 4 coordination sites / unit cell) to allow insertion of self (iron) atoms, they have excellent corrosion resistance, thermal conductivity, and nonmagnetic properties. After their melting, casting, and machining to required dimensions and geometry, stabilizing heat treatment is applied to precipitate all carbon as TiC and prevent formation of Cr23C6 (sensitization). This resist heat and surface degradation and yield excellent architecture which not only inhibit Frankel pair generation but will also allow bulk assimilation or surface annihilation (loop punching) of this lattice point defect. A superior thermal, fluid, and structural design augment above

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“…To this end, various materials (such as W [47][48][49][50], addition of Rh in W [51], use of bcc Fe [52,53], Ta [54], W-Ta [55], Ta/Fe [56], Pd [57], nanocrystalline Cu [58], SiOC/Crystalline Fe nanocomposite [59], W-K [60], reduced activation steel [61], ferritic [62], ferritic/martensitic steels [63], Be pebbles [64][65][66][67], Be and beryllides [68], graphite, carbon fiber composite [69]) and high Z atoms (Zr, No, Mo, Hf, Ta) [70] have been tested but none proved satisfactory [71][72][73][74]. All show rapid surface degradation exhibiting surface blisters [75][76][77][78] and formation of fuzz [51,[79][80][81][82] or under dense nanostructure [40] after bubble. These rapidly degrade their physical (e.g.…”

Section: Introductionmentioning

confidence: 99%

Plasma Facing Material by Self-Interstitial Solid Solution Strengthening – Problem, Proposition, and a Solution

Rafique¹

2023

Preprint

0

View full text Add to dashboard Cite

Bubble (point defect) – a precursor of fuzz or under dense nanostructure formation is crystal lattice defect. Suitable selection of crystal lattice which inhibit Frenkel pair generation and intrinsically promotes selfinterstitial solid solution strengthening contributes effectively towards making plasma facing material. For this, interstitial sites, their size, amount / fraction, positions, tendency of occupation and diffusion parameters (e.g. activation energies (Q), activation volumes) are determined. Fcc iron carbon alloys (austenitic stainless steels AISI / SAE 321, fcc structure, Pearson code cF4, space group Fm3̅m) are proposed as suitable candidates. Along with their room temperature fcc structure having 12 interstitial positions (4 octahedral, 6 coordination sites and 8 tetrahedral, 4 coordination sites / unit cell) to allow insertion of self (iron) atoms, they have excellent corrosion resistance, thermal conductivity, and nonmagnetic properties. After their melting, casting, and machining to required dimensions and geometry, stabilizing heat treatment is applied to precipitate all carbon as TiC and prevent formation of Cr23C6 (sensitization). This resist heat and surface degradation and yield excellent architecture which not only inhibit Frankel pair generation but will also allow bulk assimilation or surface annihilation (loop punching) of this lattice point defect. A superior thermal, fluid, and structural design augment above

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“…Despite the importance of the material, few experimental studies have considered its use in radiation environments, − and even fewer irradiation experiments have been performed, chiefly in the 1970s with the aim of hardening the material for cutting tools. − This is in contrast to W metal, in which the microstructural evolution under irradiation, and the corresponding changes in material properties, have been extensively characterized through experimental irradiation with He, − H isotopes, − heavy ions, ,− and neutrons. − A particular concern of W (both alloyed and unalloyed), is the segregation of Re and Os transmutation products, first in the form of clusters within the W matrix and later as Re–Os-rich precipitates, ,, which are known to aggravate the radiation-induced embrittlement of W metal. , The fluence at which these precipitates appear, and when they start to dominate the degradation of the mechanical properties, is dependent on alloy type, temperature, and neutron flux. , …”

Section: Introductionmentioning

confidence: 99%

Radiation-Induced Evolution of Tungsten Carbide in Fusion Reactors: Accommodation of Defect Clusters and Transmutation Elements

Oliver

¹

,

Jackson

²

,

Burr

³

2019

ACS Appl. Energy Mater.

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The development of suitable shielding material is a key challenge in the advancement of spherical tokamak fusion reactors. Tungsten carbide (WC) is a promising candidate material owing to its low neutron and gamma attenuation lengths resulting from the combination of high-Z and low-Z elements, but its behavior under prolonged exposure to fusion neutrons is poorly understood. Here, we shed light on the microstructural evolution of WC under neutron irradiation by investigating the formation and clustering of defects using density functional theory atomic simulations. It is found that deviation from stoichiometry is accommodated entirely by C defects (vacancies and interstitials), while the disorder induced by radiation damage may be accommodated by three competing processes (C- Frenkel, antisite, and Schottky) with similar energetics. Vacancy clusters involving a combination of both tungsten and carbon vacancies show increasingly favorable binding energies with increasing cluster size and may lead to the formation of undesirable extended defects such as dislocation loops and voids. The accommodation of the three main transmutation elements Re, Os, and Ta was also considered, as well as their interaction with radiation-induced intrinsic defects. It was found that all three species are preferentially accommodated as substitution defects on the W sublattice, and they all exhibit a thermodynamic drive to bind with W vacancies, while Ta also binds with C vacancies and with Re and Os substitutions. This suggests that radiation-induced vacancy sinks (e.g., voids and dislocation loops) will likely be enriched in transmutation elements, and that these further stabilize the vacancy clusters, potentially aggravating swelling. However, it is predicted that under equilibrium conditions, all three species exhibit limited solubility in WC, assuming that the competing phases for precipitation are TaC(s), Os(s), and Re(s). Further investigation into the C-(Ta,Re,Os)-W phase diagram is needed to improve the accuracy of these predictions.

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Irradiation hardening induced by blistering in tungsten due to low-energy high flux hydrogen plasma exposure

Cited by 20 publications

References 43 publications

Plasma Facing Material by Self-Interstitial Solid Solution Strengthening – Problem, Proposition, and a Solution

Plasma Facing Material by Self-Interstitial Solid Solution Strengthening – Problem, Proposition, and a Solution

Plasma Facing Material by Self-Interstitial Solid Solution Strengthening – Problem, Proposition, and a Solution

Radiation-Induced Evolution of Tungsten Carbide in Fusion Reactors: Accommodation of Defect Clusters and Transmutation Elements

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