Thermal desorption of helium from defects in nickel

Cao, Xingzhong; Xu, Qiu; Sato, Koichi; Yoshiie, T.

doi:10.1016/j.jnucmat.2011.02.049

Cited by 26 publications

(21 citation statements)

References 22 publications

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“…Helium diffusion was investigated in a large number of metallic materials. It includes Be [27], C [30], Mg [28], Al [29,31,32], Si [33], Ti [28,34], V [35][36][37][38][39], Fe [40][41][42][43][44][45][46][47], Ni [40,[48][49][50][51][52][53][54], Cu [55,56], Zr [40], Nb [37,56], Mo [57,58], Pd [59], Ag [25], Ta [60], W [61][62][63][64][65][66], Pt [67], Au [25], Ag/Au alloy [79], Fe-based alloys [40,…”

Section: A Class I: Experimental Investigations On Pure Metals and Amentioning

confidence: 99%

Helium mobility in advanced nuclear ceramics

Agarwal

Trocellier

Serruys

et al. 2014

Nuclear Instruments and Methods in Physics Research Section B:

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Section: A Class I: Experimental Investigations On Pure Metals and Amentioning

confidence: 99%

Helium mobility in advanced nuclear ceramics

Agarwal

Trocellier

Serruys

et al. 2014

Nuclear Instruments and Methods in Physics Research Section B:

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“…As a result of its low solubility in metals, He becomes trapped in regions of excess volume, such as dislocations, grain boundaries, and, most strongly, vacancies and vacancy clusters. [5][6][7][8][9][10][11][12] As such, it aids the nucleation, stabilization, and growth of voids (He bubbles), resulting in swelling of the material. 10,[13][14][15][16] The formation of He bubbles has also been implicated in hightemperature embrittlement of materials.…”

Section: Introductionmentioning

confidence: 99%

First-principles study of helium, carbon, and nitrogen in austenite, dilute austenitic iron alloys, and nickel

et al. 2013

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An extensive set of first-principles density functional theory calculations have been performed to study the behavior of He, C, and N solutes in austenite, dilute Fe-Cr-Ni austenitic alloys, and Ni in order to investigate their influence on the microstructural evolution of austenitic steel alloys under irradiation. The results show that austenite behaves much like other face-centered cubic metals and like Ni in particular. Strong similarities were also observed between austenite and ferrite. We find that interstitial He is most stable in the tetrahedral site and migrates with a low barrier energy of between 0.1 and 0.2 eV. It binds strongly into clusters as well as overcoordinated lattice defects and forms highly stable He-vacancy (V m He n ) clusters. Interstitial He clusters of sufficient size were shown to be unstable to self-interstitial emission and VHe n cluster formation. The binding of additional He and V to existing V m He n clusters increases with cluster size, leading to unbounded growth and He bubble formation. Clusters with n/m around 1.3 were found to be most stable with a dissociation energy of 2.8 eV for He and V release. Substitutional He migrates via the dissociative mechanism in a thermal vacancy population but can migrate via the vacancy mechanism in irradiated environments as a stable V 2 He complex. Both C and N are most stable octahedrally and exhibit migration energies in the range from 1.3 to 1.6 eV. Interactions between pairs of these solutes are either repulsive or negligible. A vacancy can stably bind up to two C or N atoms with binding energies per solute atom up to 0.4 eV for C and up to 0.6 eV for N. Calculations in Ni, however, show that this may not result in vacancy trapping as VC and VN complexes can migrate cooperatively with barrier energies comparable to the isolated vacancy. This should also lead to enhanced C and N mobility in irradiated materials and may result in solute segregation to defect sinks. Binding to larger vacancy clusters is most stable near their surface and increases with cluster size. A binding energy of 0.1 eV was observed for both C and N to a [001] self-interstitial dumbbell and is likely to increase with cluster size. On this basis, we would expect that, once mobile, Cottrell atmospheres of C and N will develop around dislocations and grain boundaries in austenitic steel alloys.

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“…The significant difference in the retention properties between the D 2 + only and He + –D 2 + samples was caused by differences in the microstructures of the samples. Only small dislocation loops were formed in the sample irradiated with D + ions, whereas both dislocation loops and bubbles were formed in the sample irradiated with He + ions3238. Helium atoms were captured by dislocations and vacancies during implantation and diffusion, thereby forming stable He-vacancy complexes.…”

Section: Resultsmentioning

confidence: 99%

Effects of zirconium element on the microstructure and deuterium retention of W–Zr/Sc2O3 composites

Chen

Luo

Zan

et al. 2016

Sci Rep

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Nuclear fusion energy is considered to be the principal way to effectively solve the future energy problem as a clean and infinite energy resource. And it is being developed internationally via the International Thermonuclear Experimental Reactor (ITER) Project, which aims to demonstrate the extended burn of deuterium-tritium (D-T) plasma in a fusion reaction 1,2 . The mechanical property of plasma facing materials (PFMs) under D-T plasma irradiation is one of the most important issues for the ITER project. Tungsten (W) and its alloys are primary candidate plasma facing materials for the divertor and the first wall in fusion power reactors because of their high melting point, high thermal conductivity, high strength at elevated temperatures, low sputtering yield in radiation environment and low tritium inventory [3][4][5] . The shortcomings of pure tungsten and its alloys, such as high temperature, embrittlement problems (e.g., low-temperature brittleness, high-temperature or recrystallization brittleness, and radiation-reduced brittleness and hardness), high ductile-to-brittle transition temperature and low recrystallization, exert a negative influence on their applications and restrict their utilization [6][7][8] . Thus, developing novel tungsten materials with improved ductility and stability against high temperatures and irradiation properties is of utmost importance.Many studies have shown that several disperse second-phases particles (e.g., ThO 2 , La 2 O 3 , CeO 2 , Y 2 O 3 , and TiC) can effectively inhibit recrystallization and grain growth as well as improve high-temperature strength and creep resistance by hindering grain boundary (GB) sliding [9][10][11][12][13] . Rare earth elements are usually doped into tungsten matrix composites to refine grains, strengthen tungsten grains, increase interface bonding, affect the distribution and morphology of impurities, and so on. The impurities existing in GBs, such as carbon (C), oxygen (O), and nitrogen (N), can seriously affect the wettability between the second-phase particles and tungsten grain/s and

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Thermal desorption of helium from defects in nickel

Cited by 26 publications

References 22 publications

Helium mobility in advanced nuclear ceramics

Helium mobility in advanced nuclear ceramics

First-principles study of helium, carbon, and nitrogen in austenite, dilute austenitic iron alloys, and nickel

Effects of zirconium element on the microstructure and deuterium retention of W–Zr/Sc2O3 composites

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