The influence of high grain boundary density on helium retention in tungsten
Introduction
Tungsten has been proposed as a convenient material for plasma-facing components (PFC) in future fusion reactors since it offers several advantages: high melting point, high thermal conductivity, low sputtering coefficient and low tritium retention. Owing to these properties, tungsten is expected to be suitable for the first-wall and the divertor in magnetic fusion reactors and the first-wall (armour) in inertial fusion confinement chambers [1], [2], [3], [4]. In both technologies the most adverse irradiation events are of pulsed nature: in the case of IFE (Inertial Fusion Energy) due to the target explosions in which the technology is based and in the case of MFE (Magnetic Fusion Energy) because of the edge-localized modes [5], which up to now (e.g., ITER) are intrinsic to TOKAMAK operation in H-mode.
In direct-target IFE, the energy will be released as follows: ∼1% in the form of X-rays with a pulse duration of a few ns, not considered to be a great threat to the wall; ∼71% due to pulsed neutron irradiation, of which energy will be absorbed beyond the first-wall; and ∼27% due to pulsed ion irradiation (burn products and debris ions) [6], [7]. The resulting intense ion pulses will cause, on the one hand, very high thermal loads and on the other hand, damage due to Frenkel-Pair (FP) production and ion retention. Provided that the first wall is situated far enough from the target explosions, an acceptable thermo-mechanical response is feasible [8], [9], [10]. As regards the damage, different ions and burn products will reach the first wall, He ions among them [11]. The energy of He ions can be as high as 4 MeV with an average value of ∼2.5 MeV after thermalization in the compressed target [12], more than enough to produce Frenkel Pairs. Since He is not soluble in tungsten, He atoms tend to nucleate inside vacancies, which could result under certain circumstances in dramatic He bubble formation [13], [14]. In general, HenVm clusters (not only large bubbles) cause microstructural changes [15], [16], that may develop into blistering, cracking and exfoliation of the material [17], [18], all of them detrimental to the armour [19]. Experimental results show that continuous He irradiation (typical flux 1010–1014 cm−2 s−1) leads to deleterious effects (adverse porosity) at fluences higher than 1017–1018 He cm−2 [15], [17]. However, Renk et al. [6] showed that pulsed He irradiation (with fluxes up to 2 × 1019 cm−2 s−1) leads to detrimental effects (pore formation and protrusions) at fluences as low as 1015 He cm−2.
Many efforts are being carried out in order to mitigate the effect of He retention. A possible solution would be the use of nanostructured tungsten due to its large grain-boundary density: As grain boundaries may act as defect sinks for vacancies, interstitials and He atoms, nanostructured tungsten is expected to present higher irradiation tolerance [20]. The defects are supposed to accumulate at grain boundaries, where vacancies and self interstitial atoms (SIAs) annihilate [21]. On the other hand, He atoms can become trapped if the binding energy is high enough and diffusivity along the grain boundaries is low. Nanocolumnar tungsten has recently been fabricated with a high grain boundary density [22] and it has been experimentally observed that a high grain boundary density has a direct influence in the retention of light species [23]. A point not elucidated yet is whether inter-grain He migration allows for efficient He release or if the distribution and size of defects in the interior of a grain is affected by the high density of grain boundaries.
In the present study, we investigate the influence of high grain-boundary density in damage production by pulsed He irradiation in tungsten, by comparing monocrystalline tungsten (MW) to nanocrystalline tungsten (NW). We have carried out OKMC (Object Kinetic Monte Carlo) simulations parameterized with the aid of DFT calculations to describe the damage distribution in both materials. MMonCa code was used for this purpose [24], [25]. In order to compare the influence of high grain-boundary density in the damage distribution in both nanocrystalline and monocrystalline tungsten, we have carried out two different sets of simulations with the experimental conditions used by Renk et al. in Ref. [6].
Section snippets
Density Functional Theory (DFT) calculations
Calculations based on DFT techniques were performed using the Vienna Ab initio Simulation Package (VASP) [26], [27], [28]. The PBE [29] parameterization of the Generalized Gradient Approximation (GGA) for the exchange and correlation functional was used as well as the Plane Augmented Wave pseudopotentials [30], provided by the code. Six valence electrons have been considered for W (4 3d and 2 4s) and two 1s valence electrons for He. Within these approximations, the lattice parameter was
Results
In view of the retained He fraction, no surface He desorption is observed in our simulations. In the case of NW, an almost constant value of retained He (∼50%) in the interior of the grain is reached after ∼100 pulses and all these He atoms are retained in HenVm clusters. The other 50% remains trapped at grain boundaries, assumed to be perfect sinks. In the case of MW, the He remains retained in HenVm clusters. The He depth profiles, Fig. 5(a) show that in both cases most He is retained in a
Discussion
He implantation at 625 keV leads to a large amount of Frenkel Pairs (38.44 FP per implanted ion), mostly in the “damaged volume” (300 × 50 × 50 nm3) mentioned before, which is located at around 1000 nm in depth. The results show clearly that: (i) the damage is in the form of mixed HenVm clusters, (ii) these mixed HenVm clusters are concentrated in a small “damaged volume”, and (iii) in the interior of a grain in NW, the He/V ratio is low and thus, low pressure is expected whereas in MW the trend is
Conclusions
We have used an OKMC model parameterized with DFT data to simulate He irradiation in tungsten up to a fluence of ∼1016 ion cm−2 and a relatively high temperature (up to 1500 K). A comparison with experimental data on He desorption has been carried out in order to validate our parameterization at these temperatures. Our results are in good agreement with the experimental ones.
Simulations of 625 keV He ion irradiation in MW and NW reveal 100% He retention and the appearance of different defect
Acknowledgements
The authors acknowledge Spanish MINECO for funding through the projects ACI-A-2011-0718 (MATFLUSA) and ENE-2012-39787-C06-03 (RADIAFUS-3) and financial support from the FP7 project RADINTERFACES. I. Martin-Bragado acknowledges funding from the “Subprograma Ramón y Cajal” by the Spanish Ministry of Economy and Competitiviness. C. Gonzalez and R. Iglesias acknowledge technical support and CPU time from Ángel Gutiérrez and the Scientific Modelling Cluster unit at UNIOVI and the support provided by
References (49)
- et al.
J. Nucl. Mater
(1998) - et al.
Fusion Eng. Des.
(2000) - et al.
J. Nucl. Mater.
(2005) - et al.
J. Nucl. Mater.
(2003) J. Nucl. Mater.
(2005)- et al.
Fusion Eng. Des.
(2006) - et al.
J. Nucl. Mater.
(2005) - et al.
Fusion Eng. Des.
(2011) - et al.
J. Nucl. Mater.
(2009) - et al.
J. Nucl. Mater.
(2012)
J. Nucl. Mater.
J. Nucl. Mater
Appl. Surf. Sci.
J. Nucl. Mater.
Comput. Phys. Commun.
J. Nucl. Mater.
J. Nucl. Mater.
Nucl. Instrum. Methods Phys. Res. Sect. B Beam Interact. Mater. At
J. Nucl. Mater.
Nucl. Instrum. Methods Phys. Res. Sect. B Beam Interact. Mater. At
J. Nucl. Mater.
J. Nucl. Mater.
J. Nucl. Mater.
Int. J. Plast.
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