- Email: [email protected]
Microstructure evolution of modified 310S austenitic stainless steels under argon ion irradiation at different temperatures
Nuclear Inst. and Methods in Physics Research B 459 (2019) 7–14
Contents lists available at ScienceDirect
Nuclear Inst. and Methods in Physics Research B journal homepage: www.elsevier.com/locate/nimb
Microstructure evolution of modified 310S austenitic stainless steels under argon ion irradiation at different temperatures
T
Yaxia Weia, Zhenyu Shena, Weiping Zhanga, Rui Tangb, Yunxiang Longa, Cheng Chena, ⁎ Xiong Zhoua, Liping Guoa, , Shui Qiub a
Hubei Nuclear Solid Physics Key Laboratory, Key Laboratory of Artificial Micro- and Nano-structures of Ministry of Education and School of Physics and Technology, Wuhan University, Wuhan, Hubei 430072, China b Science and Technology on Reactor Fuel and Materials Laboratory, Nuclear Power Institute of China, Chengdu, Sichuan 610041, China
ARTICLE INFO
ABSTRACT
Keywords: Ion irradiation Austenitic steel Vacancy clusters Irradiation-induced precipitation
The effects of oversized additive atoms on microstructure evolution in 310S stainless steel under irradiation was investigated. Ar ion irradiations were conducted on modified 310S stainless steels at 290 °C and 550 °C. SC-1 was modified with Zr and SC-2 was modified with Nb, Ta and W. Compared with SC-2, lower density, smaller vacancy clusters formed in SC-1 at 290 °C. When irradiation temperature was increased from 290 °C to 550 °C, the average size of vacancy clusters increased while the number density dropped to a lower value in SC-1; however, in SC-2, the average size and number density of vacancy clusters shifted to larger values. At 550 °C, lower density and larger size of vacancy clusters were observed in SC-1 compared with that in SC-2. In addition, precipitates enriched with Ni were observed in SC-1, while Nb- and Ta-enriched precipitates were found in SC-2. A possible mechanism is discussed.
1. Introduction Supercritical water reactors (SCWRs) are one of gen-IV reactors that have high thermal efficiency and a simple construction structure [1,2]. The structural materials of SCWRs work under an environment with irradiation of neutrons and erosion of supercritical water, which can lead to material degradation, including corrosion, oxidation, stress corrosion cracking, creep rupture, and swelling. The operating temperature of SCWR structural materials ranges from 290 °C to 600 °C and the maximum dose can be up to 30 dpa (displacement per atom) [3,4]. Austenitic stainless steels are considered to be one of the primary options as structural materials for this application due to their good creep resistance at high temperature and reasonable corrosion/oxidation resistance [5,6]. Among various austenitic steel candidates, assessments show that 310S stainless steels may meet the criteria for resistance to corrosion, oxidation, stress corrosion cracking, and creep rupture; unfortunately, void swelling still remains a problem [7,8]. Many experimental results have revealed that addition of oversized, solute elements with larger size than that of solvent elements, could improve the swelling-resistance of materials [9–16]. Simulations have been conducted to discuss the corresponding mechanism [17–19]. The presence of an elastic interaction between vacancies and oversized
⁎
elements has been confirmed [10], owing to which, the recombination of point defects is enhanced and diffusivity of vacancies is suppressed, which reduces void swelling [10,20]. As mentioned, structural materials of SCWRs are exposed to a wide temperature range. This means that a thorough understanding of the effects of oversized additive elements on microstructure evolution at different temperatures, when applied to improve the irradiation-resistance of materials, is of great importance. Research on the temperature dependence of oversized additive atoms on microstructure evolution is limited [10,15] and the mechanism is not clear. In our recent article [21], we studied the irradiation properties of two types of modified 310S austenitic stainless steels, SC-1 (modified with Zr) and SC-2 (modified with Nb, Ta, and W), by proton irradiation at 290 °C. The effects of proton irradiation to materials is very similar with that of neutron irradiation, however, the dose rate of proton irradiation is very low. Thus, the proton irradiation experiment was only performed below the damage dose of 0.3 dpa. In this paper, taking advantage of high dose rate of Ar ion irradiation, we used it to investigate and discuss irradiation properties of two modified 310S austenitic stainless steels at high damage doses (up to 30 dpa), and further studied the effect of irradiation temperature at 290 °C and 550 °C. After irradiation, the specimens were observed by transmission electron
Corresponding author. E-mail address: [email protected] (L. Guo).
https://doi.org/10.1016/j.nimb.2019.08.016 Received 15 July 2019; Received in revised form 23 August 2019; Accepted 24 August 2019 Available online 28 August 2019 0168-583X/ © 2019 Elsevier B.V. All rights reserved.
Nuclear Inst. and Methods in Physics Research B 459 (2019) 7–14
Y. Wei, et al.
Fig. 2. Transmission electron microscopy images of unirradiated materials: (a) SC-1; (b) SC-2.
3. Results 3.1. Vacancy clusters +
Fig. 1. Depth profiles of damage events at 120-keV Ar irradiations in SC-1 and SC-2.
Over- and under-focused TEM images were taken of the same area in specimens to ensure that the vacancy clusters were formed in the materials after irradiation. As shown in Fig. 3, the white dots in Fig. 3(a) turned to black in Fig. 3(b), which meant that they were vacancy clusters. The over- and under-focused distances were around ± 700 nm. Fig. 4 shows TEM images of irradiated specimens. For both SC1 and SC-2, vacancy clusters were found in all specimens except for those irradiated to 5 dpa at 290 °C. Diameters and number density of vacancy clusters were measured in each specimen. The clusters in the specimens irradiated to 5 dpa at 550 °C were all smaller than 1 nm, making the counting difficult. For this reason, no information concerning the size and number density of vacancy clusters in those samples was collected.
microscope (TEM). The effects of oversized additive atoms on microstructure evolution in 310S stainless steels at high damage doses and different temperatures were investigated. 2. Experimental The materials used in this research were provided by the Nuclear Power Institute of China. Their production procedure has been explained in the published literature [22]. The chemical compositions of the materials were detailed in Ref. [21]. The preparation procedure for the TEM specimens was described in detail in our previous study [23]. The irradiations were conducted in the ion implanter JZM5900 at the Accelerator Laboratory of Wuhan University. Ar ions with an energy of 120 keV were implanted into the specimens at temperatures of 290 ± 5 °C and 550 ± 5 °C. The depth profiles of the damage doses were calculated using SRIM-2013 software [24], using a displacement energy of 40 eV [25]. The calculation model adopted in our work was quick calculation of damage [26]. The results are shown in Fig. 1. The peak damage doses are 5 dpa corresponding to the fluences of 4.1 × 1019 m−2, 15 dpa corresponding to 1.23 × 1020 m−2 and 30 dpa corresponding to 2.46 × 1020 m−2, respectively. For clarity, all the damage doses used in this study are reported at the peak point. Despite the differences in the chemical compositions between the two materials, the depth profiles of the damage doses were almost the same. Comparisons of the SRIM results between proton irradiation in Ref. [21] and the Ar ion irradiation in this paper show the differences between these two irradiation conditions: the implantation depth of proton is deeper than that of Ar ion, however, the damage dose of Ar ion irradiation is great higher than that of proton irradiation. Before and after irradiation, specimens were examined using a JEM-2010HT TEM. Thickness of the observed areas was measured by the fringe numbers from the edge of the specimen and ranged from 80 to 120 nm [27]. Measurements of energy-dispersive X-ray spectroscopy (EDS) were conducted in a JEM-2012FEF TEM. The operating voltage was 200 kV. The spot size for EDS analysis was about 20 nm. Fig. 2 shows TEM images of the unirradiated specimens.
Fig. 3. (a) Under-focus and (b) over-focus transmission electron microscopy images of SC-1 irradiated to 30 dpa at 550 °C. 8
Nuclear Inst. and Methods in Physics Research B 459 (2019) 7–14
Y. Wei, et al.
of vacancy clusters were found in the two materials. In SC-1, as the irradiation temperature changed from 290 °C to 550 °C, the average size of vacancy clusters increased while the number density dropped to a lower value; however, in SC-2, the average size and number density of vacancy clusters shifted to larger values. At the higher temperature of 550 °C, the vacancy cluster size increased in SC-1 was more than that of SC-2, resulting in higher number density and lower average size of vacancy clusters in SC-2. Fig. 6 shows size distributions of vacancy clusters in specimens irradiated at different conditions. It is clear that at 290 °C, their sizes in SC-1 were distributed at lower values compared to SC-2. At 550 °C, the situation was reversed, i.e., the number density of small vacancy clusters (diameter of less than ~2 nm) in SC-2 was much higher than in SC1, and there were more large vacancy clusters (diameter of less than ~2 nm) in SC-1. 3.2. Precipitates No precipitate was observed after irradiation at 290 °C. In contrast, precipitates formed in all specimens irradiated at 550 °C. Fig. 7 shows the TEM images of the precipitates. The average diameters and densities of precipitates in those specimens are shown in Fig. 8. After irradiation at 550 °C, the precipitates were elongated and we measured the length of the long axis. It was found that, with the increase of irradiation dose, the average size and number density of precipitates in SC-2 increased monotonically, but, for SC-1, both the precipitate size and number density increased initially and then saturated at 15 dpa. Moreover, precipitates in SC-1 had smaller sizes and higher number density than SC-2. Energy spectra of precipitates in irradiated specimens for both materials are shown in Fig. 9. Calculated compositions of precipitates based on the energy spectra are provided in Table 1. Compared with the composition in Ref. [21], it can be observed that, after irradiation, the precipitates in SC-2 were enriched with Nb and Ta, while those in SC-1 were enriched with Ni. 3.3. Comparison with proton irradiation results In our previous study [21], we investigated the proton irradiation induced defects in SC-1 and SC-2 under the maximum dose of 0.3 dpa at the irradiation temperature of 290 °C. It was revealed that more vacancy-type defects produced in SC-2 than that in SC-1, and this trend became more obvious with the dose increasing. The mean size and number density of dislocation loops in SC-2 were slightly larger than that in SC-1. Both positron annihilation spectroscopy (PAS) and TEM observations showed that irradiation damage inSC-1was less serious than that SC-2. In present study, much higher damage doses i.e. 5 dpa, 115 dpa, and 30 dpa were induced by Ar ion irradiation at 290 °C and 550 °C. The dislocation loops were much larger than that induced under 0.3dpa in our previous study, and they entangled with each other and with dislocation lines such that they were very difficult to resolve, so the dislocation loops are not presented and discussed here. As described in Section 3.1, at the irradiation temperature of 290 °C, vacancy clusters produced by such higher dose irradiation exhibited the same trends as that by low dose irradiation. At elevated temperature of 550 °C, both average size and number density of vacancy clusters in SC-2 were greater than that at 290 °C, while cluster size increased but cluster density decreased relative to the values at 290 °C in SC-1. As a whole, the total volume of vacancy clusters in SC-2 was larger than that in SC1. These results further suggested that SC-1 has better irradiation resistance than SC-2.
Fig. 4. Transmission electron microscopy images of SC-1 and SC-2 irradiated at different conditions.
Fig. 5 gives the average diameters and densities of vacancy clusters in the specimens. Observation was made in different areas for each specimen. In each area, vacancy clusters were counted to give the number density and an average size. The results shown in Fig. 5 are from different areas, thus, the error bars in the graph are standard deviations of the data. The data shows a similar dose dependence in all specimens. For the same temperature, with a higher dose, the number density of vacancy clusters initially increased but then decreased, while the average diameter kept increasing. At 290 °C, the vacancy cluster size and number density in SC-2 were greater than that in SC-1. Moreover, different temperature dependences
4. Discussion The difference between the vacancy clusters in the two materials mainly ‘resulted from variation of composition. SC-1 was modified with 9
Nuclear Inst. and Methods in Physics Research B 459 (2019) 7–14
Y. Wei, et al.
Fig. 5. (a) Average diameter and (b) number density of vacancy clusters measured in different SC-1 and SC-2 specimens.
Fig. 6. Size distribution of vacancy clusters for SC-1 and SC-2 specimens irradiated at different conditions: (a) 15 dpa, 290 °C; (b) 30 dpa, 290 °C; (c) 15 dpa, 550 °C; (d) 30 dpa, 550 °C.
Zr at a concentration of 0.12 at%, while SC-2 was modified with Nb, W, and Ta at concentrations of 0.11 at%, 0.22 at%, and 0.10 at%, respectively. All these solute elements are oversized in austenitic steels. It has been reported that addition of oversized atoms can enhance
vacancy-interstitial recombination and reduce mobility of vacancies [10,20] These atoms also act as nucleation sites for vacancy clusters [10]. The effects are related to vacancy trapping by oversized additive atoms [10]. 10
Nuclear Inst. and Methods in Physics Research B 459 (2019) 7–14
Y. Wei, et al.
Fig. 7. Transmission electron microscopy images of precipitates in SC-1 and SC-2 specimens irradiated at 550 °C. The arrows indicate the locations of precipitates. V V CV CSol and Sol vacancy clusters, respectively; Sol Vn CV CSol Vn are capture rates of vacancies by solute atoms and complexes of solute atoms and vacancies, respectively; Vn CSol Vn is the release rate of vacancy clusters from solute atoms; RVn CSol Vn CI is the recombination rate of vacancies trapped by solute atoms with SIAs. For simplification, emission of vacancies from vacancy clusters and trapping of interstitial atoms by vacancy clusters were not taken into account. The capture coefficient of vacancies by solute atoms, recombination coefficient of vacancies trapped by solute atoms with SIAs, and release coefficient of vacancy clusters from solute atoms are, respectively, expressed as follows [19,29]:
We used a simple model on the basis of the kinetic-rate theory [28] to explain the phenomena in the present research. The time dependences of the concentrations of vacancies, vacancy clusters, and solute–vacancy complexes follow Eqs. (1)–(3), respectively:
dCV dt = GV
KIV CI CV
CSol dCVn dt dCSol dt
= Vn
n=1
V 2 V CV
V Sol Vn CV CSol Vn
V Vn 1 CV CVn 1
=
V Vn CV CVn
V Sol Vn 1 CV CSol Vn 1
CSol
n=2
Vn CI
+ RVn +1 CSol
V Vn CV CVn
+
V CSol V
+
Vn CSol Vn
V Sol Vn CV CSol V Vn + 1 CI
V Sol CV
(1) (2) Vn CSol Vn
V Sol Vn
= 4 rV DV ND
(5)
RVn = 4 ri Di ND
RVn (3)
Vn
where Cx represents the concentration of defects or defect clusters. GV is the production rate of vacancies; xy represents the reaction coefficient between defect x and defect y; Vn and RVn are release and recombination coefficients of vacancies in solute atoms, respectively; KIV CI CV is the recombination rate of vacancies and self-interstitial atoms (SIAs); V 2 V V CV and Vn C CVn are the capture rates of vacancies by vacancies and
=
DVn a02
exp
(4)
EVbn kB T
(6)
where rV and rV represent capture and recombination radii, respectively; Dx is the diffusion coefficient of defect or defect cluster x; ND is alloy number density; kB is the Boltzmann constant; T is temperature; E bVn is binding energy of Vn and solution atoms. According to Eq. (1), vacancies produced by irradiation follow three different evolution
V
Fig. 8. (a) Average diameter and (b) number density of precipitates formed at SC-1 and SC-2 specimens after irradiation at 550 °C. 11
Nuclear Inst. and Methods in Physics Research B 459 (2019) 7–14
Y. Wei, et al.
Fig. 9. Energy spectra of precipitates in specimens: (a) SC-1 irradiated to 15 dpa at 550 °C; (b) SC-2 irradiated to 15 dpa at 550 °C.
Table 1 Compositions (wt%) of precipitates calculated based on EDS spectroscopy for SC-1 (modified with Zr) and SC-2 (modified with Nb, Ta, and W).
S P Si Mn Cr Ni Mo Nb W V Ta Zr Fe
SC-1 Precipitates
SC-2 Precipitates
0.7 0.6 0.8 0.0 20.2 24.9 0.0 – – 0.2 – 0 52.7
5.1 0.0 11.4 0.0 9.8 2.4 1.1 18.4 0.0 0.2 49.9 – 1.4
reference, the difference in recombination rate arising from the binding energy of a vacancy and solute atoms could be as much as two orders of magnitude [29]. It is quite possible that the recombination rate in SC-1 was higher than in SC-2, although the concentration of solute atoms in SC-1 was lower. We thus conclude that at 290 °C, compared with SC-1, a higher release rate and lower recombination rate in SC-2 led to the formation of vacancy clusters with high number density and great sizes, as observed in the present study. According to Eq. (6), the release rate coefficient decreases as the vacancy binding energy of additive atoms increases. It means that additive atoms with smaller binding energy with vacancies act as more effective vacancy cluster nucleation sites and less effective recombination sites. Since vacancy binding energy of Zr is higher than those of three other elements, the ratio of release rate and recombination rate at additive atoms in SC-2 is larger than that in SC-1. The concentration of additive atoms in SC-2 is four times higher than SC-1. It means that capture rate of vacancies by additive atoms in SC-2 is larger than SC-1 according to Eq. (1). It suggests that release rate of vacancy clusters from additive atoms in SC-2 is higher than that in SC-1, which means larger nucleation rate in SC-2. It is consistent to some simulation results in literatures (162.164.168). On the other hand, Zr atoms were more effective recombination sites. It is quite possible that point defect recombination rate in SC-1 is larger than that in SC-2. It means that concentration of vacancies in SC-2 is higher. It is helpful for the formation of vacancy clusters with large sizes and high densities. As the irradiation temperature was increased to 550 °C, SC-1 and SC-2 exhibited different behaviors in response to the temperature change. For SC-1, number density of vacancy clusters with size below about 1.5 nm greatly decreased and vacancy clusters above 2 nm and 3 nm could be observed in specimen irradiated to 15 dpa and 30 dpa respectively. These were not found at 290 °C. Based on Eq. (6), the release coefficient becomes large for a high irradiation temperature. Calculation of 316L with 0.1 at% Zr showed that, as the temperature increased, release rate of vacancies from solution atoms increased, but the recombination rate for vacancies and SIAs at additive sites stayed almost constant [29]. The material used in this calculation had a similar concentration of Zr to that in SC-1 [29]. As is known, recombination processes are dominant below 400 °C in our materials, while release becomes the major mechanism for temperatures above 400 °C [29]. Therefore, in SC-1, the release rate was much higher than recombination rate at 550 °C, and Zr acted mainly as a nucleation site rather than a recombination site. It is also noted that recombination could still be enhanced by Zr atoms, and swelling was thus suppressed when compared with 310S with no additive atoms. DV is also large at high temperatures, which, according to Eq. (4), results in a high capture rate of vacancies by solute atoms. Thus, a large value of capture rate of
paths; besides recombination with SIAs, some vacancies combine with vacancies and vacancy clusters, resulting in nucleation and growth of vacancy clusters, while the remaining vacancies are captured by additive atoms. Eq. (2) implies that vacancy clusters originate from coalescence of migrating vacancies or release of solute atoms. The latter mechanism reflects the enhancement of vacancy cluster nucleation by solute atoms. Eq. (3) suggests that vacancies trapped by additive atoms fall undergo two fates: recombination with interstitial atoms or release from additive atoms as vacancy-clusters. In other words, the solute atoms simultaneously act as both recombination sinks and vacancy cluster nucleation sites. Compared with SC-1, higher number density larger vacancy clusters were observed in SC-2 after 290 °C irradiation. This difference resulted from the variation of composition and concentration of additive atoms. As pointed out in the literature [10], vacancy trapping of oversized solute atoms is due to the size effect of solute atoms, which means that the higher the size factor, the larger is the binding energy of solute atoms with vacancies and vacancy clusters. Based on these literatures [10,30], it is expected that the linear size factors of additive elements rank in the order of lsfZr > lsfTa > lsfNb > lsfW . Here lsfx represents the linear size factor of element x. This implies that, of the four additive elements, Zr has the largest binding energy with vacancies and vacancy clusters [12,29]. Because the concentration of solute atoms in SC-2 was nearly four times higher than in SC-1, hence improving releasing, the release rate in SC-2 was higher than in SC-1, resulting in more effective nucleation of vacancy clusters in SC-2. Some calculation results indicate a small nucleation rate of the clusters for a higher binding energy between additive element and vacancy [18,31]. According to this 12
Nuclear Inst. and Methods in Physics Research B 459 (2019) 7–14
Y. Wei, et al.
vacancies by solute atoms in SC-1 was expected when the temperature increased from 290 °C to 550 °C. High capture rate leads to a low value of CV , which means that nucleation of vacancy clusters is less effective. Calculations in the literature [17] also show that the nucleation rate of vacancy clusters is small at high temperatures for alloys with solute additives, which can explain the observed low number density of small vacancy clusters in SC-1 at 550 °C. Considering the strong interactions between vacancies and Zr atoms, it is quite possible that there are more vacancies in vacancy clusters released from solute atoms. In other words, Zr atoms act not only as nucleation sites for vacancy clusters, but also promote their growth. This accounts for the formation of large vacancy clusters in the specimens. In addition, at a high temperature, the diffusion of vacancy clusters is enhanced. As a result, migration and coalescence of vacancy clusters are more effective, resulting in low number density and large sizes of them. Compared with SC-1, different behavior was observed in SC-2. As Fig. 5 shows, the average size and number density of vacancy clusters in SC-2 both increased slightly after irradiation at 550 °C. As mentioned earlier, Nb and Ta-enriched precipitates were observed in SC-2 irradiated at 550 °C. These caused a reduction in concentration of solute atoms. The decrease of CSol certainly could have reduced the capture rate of vacancies by solute atoms, suppressing nucleation of vacancy clusters and causing a low number density of them. However, recombination of vacancies and SIAs at additive atom sites was also simultaneously suppressed, contributing to a high value of CV . At the same time, a low capture rate also means a high value of CV . Higher temperatures and fewer solute atoms imply higher diffusivity of vacancies and vacancy clusters, leading to larger values of VV and VVn . The combined increases in CV , V Vn C V CVn .
and
Two types of 310S austenitic steels modified by different oversized elements were irradiated with Ar+ ions to damage dose up to 30 dpa at 290 °C and 550 °C respectively. On the bases of the results obtained by SRIM calculations, the kinetic-rate theory models and experimental measurements by TEM technique, the main conclusions are summarized as follows: (1) At 290 °C, Ta, Nb, and W atoms acted as effective nucleation sites for vacancy clusters, but inefficient recombination sites for vacancies and interstitial atoms, giving rise to high number density large vacancy clusters in SC-2; (2) On increasing the temperature from 290 °C to 550 °C, Zr atoms became nucleation sites for vacancy clusters instead of recombination sites for vacancies and SIAs. The capture rate of vacancies by solute atoms increased. At the same time, we believe that Zr atoms enhanced the growth of vacancy clusters because of their strong interaction with vacancies. These mechanisms explain the greatly increased sizes and low number density of vacancy clusters in SC-1 at 550 °C. (3) In SC-2, precipitation of Nb and Ta suppressed the capturing of vacancies from additive atoms and enhanced the diffusivity of vacancies, leading to enhancement of nucleation and growth of vacancy cluster, resulting in slightly higher number density and larger size of them at 550 °C. (4) Precipitates enriched with Nb and Ta were observed in SC-2 after irradiation at 550 °C due to their higher diffusivities than W and Ni. In SC-1, Zr remained dissolved in the matrix even after irradiation at 550 °C, because of its low diffusivity.
resulted in large values of In other words, there was a more effective nuand cleation process and vacancy clusters grew. Not only do these compensate for the consequence of a decreasing nucleation rate, but also contribute to the formation of high number density, large size vacancy clusters, as observed in the specimens. Also, similar to SC-1, migration and coalescence of small vacancy clusters were enhanced by high temperature in SC-2, and large ones formed. In conclusion, the different features of the vacancy clusters in the two materials after 550 °C irradiation mainly resulted from two factors: the enhanced vacancy cluster growth by Zr atoms at high temperature in SC-1 and the precipitation of Nb and Ta in SC-2. As pointed out earlier, the precipitates formed in the two materials after 550 °C irradiation were different. Ni-enriched precipitates were found in SC-1, while Ta and Nb-enriched precipitates were observed in SC-2. Precipitation of W and Zr with concentrations comparable with those of Ta or Nb did not appear in our specimens. As is well known, the growth of precipitates is closely related to the diffusion of solute atoms. It has been reported that diffusivities of Ta and Nb are higher than that of W in body-centered cubic (bcc) Fe at 1050 K [32,33]. In Ref. [34], diffusivity of Zr in bcc-Fe was calculated as being almost the same as that of Fe self-diffusivity at 1050 K [35]. This value is lower than the diffusivities of Ni, Nb and Ta at the same condition. A similar mechanism may also exist in present experimental situation; namely, having higher diffusivity than Zr, Ni precipitated in SC-1; in contrast, in SC-2, the diffusivities of Nb and Ta were both higher than those of W and Ni, leading to Nb- and Ta-enriched precipitates. In present study, Ar ions were employed to create displacement damage. Ar atoms likely remained in the foil as interstitials, which somehow played a role to trap vacancies. This implies that Ar ion irradiation induced diffusion of solute atoms will be less than that induced by self ion (i.e. Fe ions for 310S steels in present study) irradiation or neutron irradiation. In other words, less precipitates might form compared to self ions and neutrons. However, this does not influence the comparison of the precipitation behavior of two modified 310S steels (SC-1 and SC-2), because the irradiation conditions are the same. V 2 V CV
V V ,
5. Summary
V Vn
6. Data availability The raw data required to reproduce these findings cannot be shared at this time as the data also forms part of an ongoing study. The processed data required to reproduce these findings cannot be shared at this time as the data also forms part of an ongoing study. Declaration of Competing Interest The authors declare that they have no known competing financial interests or personal relationships that could have appeared to influence the work reported in this paper. Acknowledgements Financial supports from the National Natural Science Foundation of China (11775162, 11975170 and U1532134) and the International Science & Technology Cooperation Program of China (2015DFR60370) are gratefully acknowledged. The authors would like to thank the Center for Electron Microscopy at Wuhan University, China. We thank Prof. Congxiao Liu, Alabama A&M University, for the technical assistance during the preparation of this manuscript. We acknowledge the assistance of TEM specimen preparation from Center for Nanoscience and Nanotechnology at Wuhan University. Appendix A. Supplementary data Supplementary data to this article can be found online at https:// doi.org/10.1016/j.nimb.2019.08.016. References [1] K.H. Chang, S.M. Chen, T.K. Yeh, J.J. Kai, Effect of dissolved oxygen content on the oxide structure of Alloy 625 in supercritical water environments at 700 °C, Corros. Sci. 81 (2014) 21–26.
13
Nuclear Inst. and Methods in Physics Research B 459 (2019) 7–14
Y. Wei, et al. [2] Y. Behnamian, A. Mostafaei, A. Kohandehghan, B.S. Amirkhiz, D. Serate, W. Zheng, D. Guzonas, M. Chmielus, W. Chen, J.L. Luo, Characterization of oxide scales grown on alloy 310S stainless steel after long term exposure to supercritical water at 500 °C, Mater. Charact. 120 (2016) 273–284. [3] S.J. Zinkle, G.S. Was, Materials challenges in nuclear energy, Acta Mater. 61 (2013) 735–758. [4] D. Guzonas, R. Novotny, Supercritical water-cooled reactor materials – summary of research and open issues, Prog. Nucl. Energy. 77 (2014) 361–372. [5] K.L. Murty, I. Charit, Structural materials for Gen-IV nuclear reactors: challenges and opportunities, J. Nucl. Mater. 383 (2008) 189–195. [6] C. Sun, R. Hui, W. Qu, S. Yick, Progress in corrosion resistant materials for supercritical water reactors, Corros. Sci. 51 (2009) 2508–2523. [7] W. Zheng, D. Guzonas, K.P. Boyle, J. Li, S. Xu, Materials assessment for the Canadian SCWR core concept, JOM 68 (2016) 456–462. [8] F.A. Garner, Recent insights on the swelling and creep of irradiated austenitic alloys, J. Nucl. Mater. 122–123 (1984) 459–471. [9] M.J. Hackett, J.T. Busby, G.S. Was, The mechanism of Zr and Hf in reducing radiation-induced segregation in 316 stainless steel, Metall. Mater. Trans. A: Phys. Metall. Mater. Sci. 39 (2008) 218–224. [10] T. Kato, H. Takahashi, I. Izumiya, Effects of systematic modification with oversized elements on void formation in 316L austenitic stainless steel under electron irradiation, Mater. Trans. JIM 32 (1991) 921–930. [11] F.A. Garner, H.R. Brager, The influence of Mo, Si, P, C, Ti, Cr, Zr and various trace elements on the neutron-induced swelling of AISI 316 stainless steel, J. Nucl. Mater. 155–157 (1988) 833–837. [12] T. Kato, H. Takahashi, M. Izumiya, Grain boundary segregation under electron irradiation in austenitic stainless steels modified with oversized elements, J. Nucl. Mater. 189 (1992) 167–174. [13] J. Gan, E.P. Simonen, S.M. Bruemmer, L. Fournier, B.H. Sencer, G.S. Was, The effect of oversized solute additions on the microstructure of 316SS irradiated with 5 MeV Ni++ions or 3.2 MeV protons, J. Nucl. Mater. 325 (2004) 94–106. [14] Y. Katano, K. Nakata, S. Jitsukawa, T. Aruga, K. Shiraishi, Effects of titanium and zirconium on void swelling in electron-irradiated austenitic stainless steel, J. Nucl. Mater. 133–134 (1985) 530–534. [15] I. Shibahara, N. Akasaka, S. Onose, H. Okada, S. Ukai, Swellling of advanced austenitic stainless steels developed for the environment of heavy neutron exposure, J. Nucl. Mater. 212–215 (1994) 487–491. [16] F.A. Garner, H.R. Brager, R.J. Puigh, Swelling behavior of titanium-modified alloys in EBR-II, J. Nucl. Mater. 133–134 (1985) 535–539. [17] F.A. Garner, W.G. Wolfer, The effect of solute additions on void nucleation, J. Nucl. Mater. 102 (1981) 143–150. [18] L.K. Mansur, M.H. Yoo, The effects of impurity trapping on irradiation-induced swelling and creep, J. Nucl. Mater. 74 (1978) 228–241. [19] L.K. Mansur, Void swelling in metals and alloys under irradiation: an assessment of
the theory, Nucl. Technol. 40 (1978) 5–34. [20] Y. Sekio, S. Yamashita, N. Sakaguchi, H. Takahashi, Effect of additional minor elements on accumulation behavior of point defects under electron irradiation in austenitic stainless steels, Mater. Trans. 55 (2014) 438–442. [21] W.P. Zhang, Z.Y. Shen, R. Tang, S.X. Jin, Y.X. Song, Y.X. Long, Y.X. Wei, X. Zhou, C. Chen, L.P. Guo, Proton-irradiation induced defects in modified 310S steels characterized with positron annihilation spectroscopy and transmission electron microscopy, Nucl. Instrum. Methods Phys. Res. Sect. B 427 (2018) 1–8. [22] D. Wen, B. Jiang, Q. Wang, F. Yu, X. Li, R. Tang, R. Zhang, G. Chen, C. Dong, Influences of Mo/Zr minor-alloying on the phase precipitation behavior in modified 310S austenitic stainless steels at high temperatures, Mater. Des. 128 (2017) 34–46. [23] Z. Zheng, N. Gao, R. Tang, Y. Yu, W. Zhang, Z. Shen, Y. Long, Y. Wei, L. Guo, Evolution of dislocation loops in austenitic stainless steels implanted with high concentration of hydrogen, Philos. Mag. 97 (2017) 2658–2668. [24] J.F. Ziegler, M.D. Ziegler, J.P. Biersack, SRIM – The stopping and range of ions in matter, Nucl. Instrum. Methods Phys. Res. Sect. B: Beam Interact. Mater. Atoms 268 (2010) 1818–1823. [25] ASTME521, Standard Practice for Neutron Radiation Damage Simulation by Charged Particle Irradiation, in: Annu. B. ASTM Stand. Am. Soc. Test. Mater., American Society for Testing and Materials (ASTM), Philadelphia, PA, 1989: p. D-9. [26] R.E. Stoller, M.B. Toloczko, G.S. Was, A.G. Certain, S. Dwaraknath, F.A. Garner, On the use of SRIM for computing radiation damage exposure, Nucl. Instrum. Methods Phys. Res. Sect. B: Beam Interact. Mater. Atoms 310 (2013) 75–80. [27] D.B. Williams, C.B. Carter, Transmission Electron Microscopy, Springer, US, Boston, MA, 2009. [28] T. Jourdan, G. Bencteux, G. Adjanor, Efficient simulation of kinetics of radiation induced defects: a cluster dynamics approach, J. Nucl. Mater. 444 (2014) 298–313. [29] M.J. Hackett, R. Najafabadi, G.S. Was, Modeling solute-vacancy trapping at oversized solutes and its effect on radiation-induced segregation in Fe-Cr-Ni alloys, J. Nucl. Mater. 389 (2009) 279–287. [30] H.W. King, Quantitative size-factors for metallic solid solutions, J. Mater. Sci. 1 (1966) 79–90. [31] L.K. Mansur, Effects of point defect trapping and solute segregation on irradiationinduced swelling and creep, J. Nucl. Mater. 83 (1979) 109–127. [32] S. Huang, D.L. Worthington, M. Asta, V. Ozolins, G. Ghosh, P.K. Liaw, Calculation of impurity diffusivities in α-Fe using first-principles methods, Acta Mater. 58 (2010) 1982–1993. [33] H. Ding, S. Huang, G. Ghosh, P.K. Liaw, M. Asta, A computational study of impurity diffusivities for 5d transition metal solutes in α-Fe, Scr. Mater. 67 (2012) 732–735. [34] D. Murali, B.K. Panigrahi, M.C. Valsakumar, C.S. Sundar, Diffusion of y and Ti/Zr in bcc iron: A first principles study, J. Nucl. Mater. 419 (2011) 208–212. [35] A. Seeger, Lattice vacancies in high-purity α-iron, Phys. Status Solidi Appl. Res. 167 (1998) 289–311.
14
Contents lists available at ScienceDirect
Nuclear Inst. and Methods in Physics Research B journal homepage: www.elsevier.com/locate/nimb
Microstructure evolution of modified 310S austenitic stainless steels under argon ion irradiation at different temperatures
T
Yaxia Weia, Zhenyu Shena, Weiping Zhanga, Rui Tangb, Yunxiang Longa, Cheng Chena, ⁎ Xiong Zhoua, Liping Guoa, , Shui Qiub a
Hubei Nuclear Solid Physics Key Laboratory, Key Laboratory of Artificial Micro- and Nano-structures of Ministry of Education and School of Physics and Technology, Wuhan University, Wuhan, Hubei 430072, China b Science and Technology on Reactor Fuel and Materials Laboratory, Nuclear Power Institute of China, Chengdu, Sichuan 610041, China
ARTICLE INFO
ABSTRACT
Keywords: Ion irradiation Austenitic steel Vacancy clusters Irradiation-induced precipitation
The effects of oversized additive atoms on microstructure evolution in 310S stainless steel under irradiation was investigated. Ar ion irradiations were conducted on modified 310S stainless steels at 290 °C and 550 °C. SC-1 was modified with Zr and SC-2 was modified with Nb, Ta and W. Compared with SC-2, lower density, smaller vacancy clusters formed in SC-1 at 290 °C. When irradiation temperature was increased from 290 °C to 550 °C, the average size of vacancy clusters increased while the number density dropped to a lower value in SC-1; however, in SC-2, the average size and number density of vacancy clusters shifted to larger values. At 550 °C, lower density and larger size of vacancy clusters were observed in SC-1 compared with that in SC-2. In addition, precipitates enriched with Ni were observed in SC-1, while Nb- and Ta-enriched precipitates were found in SC-2. A possible mechanism is discussed.
1. Introduction Supercritical water reactors (SCWRs) are one of gen-IV reactors that have high thermal efficiency and a simple construction structure [1,2]. The structural materials of SCWRs work under an environment with irradiation of neutrons and erosion of supercritical water, which can lead to material degradation, including corrosion, oxidation, stress corrosion cracking, creep rupture, and swelling. The operating temperature of SCWR structural materials ranges from 290 °C to 600 °C and the maximum dose can be up to 30 dpa (displacement per atom) [3,4]. Austenitic stainless steels are considered to be one of the primary options as structural materials for this application due to their good creep resistance at high temperature and reasonable corrosion/oxidation resistance [5,6]. Among various austenitic steel candidates, assessments show that 310S stainless steels may meet the criteria for resistance to corrosion, oxidation, stress corrosion cracking, and creep rupture; unfortunately, void swelling still remains a problem [7,8]. Many experimental results have revealed that addition of oversized, solute elements with larger size than that of solvent elements, could improve the swelling-resistance of materials [9–16]. Simulations have been conducted to discuss the corresponding mechanism [17–19]. The presence of an elastic interaction between vacancies and oversized
⁎
elements has been confirmed [10], owing to which, the recombination of point defects is enhanced and diffusivity of vacancies is suppressed, which reduces void swelling [10,20]. As mentioned, structural materials of SCWRs are exposed to a wide temperature range. This means that a thorough understanding of the effects of oversized additive elements on microstructure evolution at different temperatures, when applied to improve the irradiation-resistance of materials, is of great importance. Research on the temperature dependence of oversized additive atoms on microstructure evolution is limited [10,15] and the mechanism is not clear. In our recent article [21], we studied the irradiation properties of two types of modified 310S austenitic stainless steels, SC-1 (modified with Zr) and SC-2 (modified with Nb, Ta, and W), by proton irradiation at 290 °C. The effects of proton irradiation to materials is very similar with that of neutron irradiation, however, the dose rate of proton irradiation is very low. Thus, the proton irradiation experiment was only performed below the damage dose of 0.3 dpa. In this paper, taking advantage of high dose rate of Ar ion irradiation, we used it to investigate and discuss irradiation properties of two modified 310S austenitic stainless steels at high damage doses (up to 30 dpa), and further studied the effect of irradiation temperature at 290 °C and 550 °C. After irradiation, the specimens were observed by transmission electron
Corresponding author. E-mail address: [email protected] (L. Guo).
https://doi.org/10.1016/j.nimb.2019.08.016 Received 15 July 2019; Received in revised form 23 August 2019; Accepted 24 August 2019 Available online 28 August 2019 0168-583X/ © 2019 Elsevier B.V. All rights reserved.
Nuclear Inst. and Methods in Physics Research B 459 (2019) 7–14
Y. Wei, et al.
Fig. 2. Transmission electron microscopy images of unirradiated materials: (a) SC-1; (b) SC-2.
3. Results 3.1. Vacancy clusters +
Fig. 1. Depth profiles of damage events at 120-keV Ar irradiations in SC-1 and SC-2.
Over- and under-focused TEM images were taken of the same area in specimens to ensure that the vacancy clusters were formed in the materials after irradiation. As shown in Fig. 3, the white dots in Fig. 3(a) turned to black in Fig. 3(b), which meant that they were vacancy clusters. The over- and under-focused distances were around ± 700 nm. Fig. 4 shows TEM images of irradiated specimens. For both SC1 and SC-2, vacancy clusters were found in all specimens except for those irradiated to 5 dpa at 290 °C. Diameters and number density of vacancy clusters were measured in each specimen. The clusters in the specimens irradiated to 5 dpa at 550 °C were all smaller than 1 nm, making the counting difficult. For this reason, no information concerning the size and number density of vacancy clusters in those samples was collected.
microscope (TEM). The effects of oversized additive atoms on microstructure evolution in 310S stainless steels at high damage doses and different temperatures were investigated. 2. Experimental The materials used in this research were provided by the Nuclear Power Institute of China. Their production procedure has been explained in the published literature [22]. The chemical compositions of the materials were detailed in Ref. [21]. The preparation procedure for the TEM specimens was described in detail in our previous study [23]. The irradiations were conducted in the ion implanter JZM5900 at the Accelerator Laboratory of Wuhan University. Ar ions with an energy of 120 keV were implanted into the specimens at temperatures of 290 ± 5 °C and 550 ± 5 °C. The depth profiles of the damage doses were calculated using SRIM-2013 software [24], using a displacement energy of 40 eV [25]. The calculation model adopted in our work was quick calculation of damage [26]. The results are shown in Fig. 1. The peak damage doses are 5 dpa corresponding to the fluences of 4.1 × 1019 m−2, 15 dpa corresponding to 1.23 × 1020 m−2 and 30 dpa corresponding to 2.46 × 1020 m−2, respectively. For clarity, all the damage doses used in this study are reported at the peak point. Despite the differences in the chemical compositions between the two materials, the depth profiles of the damage doses were almost the same. Comparisons of the SRIM results between proton irradiation in Ref. [21] and the Ar ion irradiation in this paper show the differences between these two irradiation conditions: the implantation depth of proton is deeper than that of Ar ion, however, the damage dose of Ar ion irradiation is great higher than that of proton irradiation. Before and after irradiation, specimens were examined using a JEM-2010HT TEM. Thickness of the observed areas was measured by the fringe numbers from the edge of the specimen and ranged from 80 to 120 nm [27]. Measurements of energy-dispersive X-ray spectroscopy (EDS) were conducted in a JEM-2012FEF TEM. The operating voltage was 200 kV. The spot size for EDS analysis was about 20 nm. Fig. 2 shows TEM images of the unirradiated specimens.
Fig. 3. (a) Under-focus and (b) over-focus transmission electron microscopy images of SC-1 irradiated to 30 dpa at 550 °C. 8
Nuclear Inst. and Methods in Physics Research B 459 (2019) 7–14
Y. Wei, et al.
of vacancy clusters were found in the two materials. In SC-1, as the irradiation temperature changed from 290 °C to 550 °C, the average size of vacancy clusters increased while the number density dropped to a lower value; however, in SC-2, the average size and number density of vacancy clusters shifted to larger values. At the higher temperature of 550 °C, the vacancy cluster size increased in SC-1 was more than that of SC-2, resulting in higher number density and lower average size of vacancy clusters in SC-2. Fig. 6 shows size distributions of vacancy clusters in specimens irradiated at different conditions. It is clear that at 290 °C, their sizes in SC-1 were distributed at lower values compared to SC-2. At 550 °C, the situation was reversed, i.e., the number density of small vacancy clusters (diameter of less than ~2 nm) in SC-2 was much higher than in SC1, and there were more large vacancy clusters (diameter of less than ~2 nm) in SC-1. 3.2. Precipitates No precipitate was observed after irradiation at 290 °C. In contrast, precipitates formed in all specimens irradiated at 550 °C. Fig. 7 shows the TEM images of the precipitates. The average diameters and densities of precipitates in those specimens are shown in Fig. 8. After irradiation at 550 °C, the precipitates were elongated and we measured the length of the long axis. It was found that, with the increase of irradiation dose, the average size and number density of precipitates in SC-2 increased monotonically, but, for SC-1, both the precipitate size and number density increased initially and then saturated at 15 dpa. Moreover, precipitates in SC-1 had smaller sizes and higher number density than SC-2. Energy spectra of precipitates in irradiated specimens for both materials are shown in Fig. 9. Calculated compositions of precipitates based on the energy spectra are provided in Table 1. Compared with the composition in Ref. [21], it can be observed that, after irradiation, the precipitates in SC-2 were enriched with Nb and Ta, while those in SC-1 were enriched with Ni. 3.3. Comparison with proton irradiation results In our previous study [21], we investigated the proton irradiation induced defects in SC-1 and SC-2 under the maximum dose of 0.3 dpa at the irradiation temperature of 290 °C. It was revealed that more vacancy-type defects produced in SC-2 than that in SC-1, and this trend became more obvious with the dose increasing. The mean size and number density of dislocation loops in SC-2 were slightly larger than that in SC-1. Both positron annihilation spectroscopy (PAS) and TEM observations showed that irradiation damage inSC-1was less serious than that SC-2. In present study, much higher damage doses i.e. 5 dpa, 115 dpa, and 30 dpa were induced by Ar ion irradiation at 290 °C and 550 °C. The dislocation loops were much larger than that induced under 0.3dpa in our previous study, and they entangled with each other and with dislocation lines such that they were very difficult to resolve, so the dislocation loops are not presented and discussed here. As described in Section 3.1, at the irradiation temperature of 290 °C, vacancy clusters produced by such higher dose irradiation exhibited the same trends as that by low dose irradiation. At elevated temperature of 550 °C, both average size and number density of vacancy clusters in SC-2 were greater than that at 290 °C, while cluster size increased but cluster density decreased relative to the values at 290 °C in SC-1. As a whole, the total volume of vacancy clusters in SC-2 was larger than that in SC1. These results further suggested that SC-1 has better irradiation resistance than SC-2.
Fig. 4. Transmission electron microscopy images of SC-1 and SC-2 irradiated at different conditions.
Fig. 5 gives the average diameters and densities of vacancy clusters in the specimens. Observation was made in different areas for each specimen. In each area, vacancy clusters were counted to give the number density and an average size. The results shown in Fig. 5 are from different areas, thus, the error bars in the graph are standard deviations of the data. The data shows a similar dose dependence in all specimens. For the same temperature, with a higher dose, the number density of vacancy clusters initially increased but then decreased, while the average diameter kept increasing. At 290 °C, the vacancy cluster size and number density in SC-2 were greater than that in SC-1. Moreover, different temperature dependences
4. Discussion The difference between the vacancy clusters in the two materials mainly ‘resulted from variation of composition. SC-1 was modified with 9
Nuclear Inst. and Methods in Physics Research B 459 (2019) 7–14
Y. Wei, et al.
Fig. 5. (a) Average diameter and (b) number density of vacancy clusters measured in different SC-1 and SC-2 specimens.
Fig. 6. Size distribution of vacancy clusters for SC-1 and SC-2 specimens irradiated at different conditions: (a) 15 dpa, 290 °C; (b) 30 dpa, 290 °C; (c) 15 dpa, 550 °C; (d) 30 dpa, 550 °C.
Zr at a concentration of 0.12 at%, while SC-2 was modified with Nb, W, and Ta at concentrations of 0.11 at%, 0.22 at%, and 0.10 at%, respectively. All these solute elements are oversized in austenitic steels. It has been reported that addition of oversized atoms can enhance
vacancy-interstitial recombination and reduce mobility of vacancies [10,20] These atoms also act as nucleation sites for vacancy clusters [10]. The effects are related to vacancy trapping by oversized additive atoms [10]. 10
Nuclear Inst. and Methods in Physics Research B 459 (2019) 7–14
Y. Wei, et al.
Fig. 7. Transmission electron microscopy images of precipitates in SC-1 and SC-2 specimens irradiated at 550 °C. The arrows indicate the locations of precipitates. V V CV CSol and Sol vacancy clusters, respectively; Sol Vn CV CSol Vn are capture rates of vacancies by solute atoms and complexes of solute atoms and vacancies, respectively; Vn CSol Vn is the release rate of vacancy clusters from solute atoms; RVn CSol Vn CI is the recombination rate of vacancies trapped by solute atoms with SIAs. For simplification, emission of vacancies from vacancy clusters and trapping of interstitial atoms by vacancy clusters were not taken into account. The capture coefficient of vacancies by solute atoms, recombination coefficient of vacancies trapped by solute atoms with SIAs, and release coefficient of vacancy clusters from solute atoms are, respectively, expressed as follows [19,29]:
We used a simple model on the basis of the kinetic-rate theory [28] to explain the phenomena in the present research. The time dependences of the concentrations of vacancies, vacancy clusters, and solute–vacancy complexes follow Eqs. (1)–(3), respectively:
dCV dt = GV
KIV CI CV
CSol dCVn dt dCSol dt
= Vn
n=1
V 2 V CV
V Sol Vn CV CSol Vn
V Vn 1 CV CVn 1
=
V Vn CV CVn
V Sol Vn 1 CV CSol Vn 1
CSol
n=2
Vn CI
+ RVn +1 CSol
V Vn CV CVn
+
V CSol V
+
Vn CSol Vn
V Sol Vn CV CSol V Vn + 1 CI
V Sol CV
(1) (2) Vn CSol Vn
V Sol Vn
= 4 rV DV ND
(5)
RVn = 4 ri Di ND
RVn (3)
Vn
where Cx represents the concentration of defects or defect clusters. GV is the production rate of vacancies; xy represents the reaction coefficient between defect x and defect y; Vn and RVn are release and recombination coefficients of vacancies in solute atoms, respectively; KIV CI CV is the recombination rate of vacancies and self-interstitial atoms (SIAs); V 2 V V CV and Vn C CVn are the capture rates of vacancies by vacancies and
=
DVn a02
exp
(4)
EVbn kB T
(6)
where rV and rV represent capture and recombination radii, respectively; Dx is the diffusion coefficient of defect or defect cluster x; ND is alloy number density; kB is the Boltzmann constant; T is temperature; E bVn is binding energy of Vn and solution atoms. According to Eq. (1), vacancies produced by irradiation follow three different evolution
V
Fig. 8. (a) Average diameter and (b) number density of precipitates formed at SC-1 and SC-2 specimens after irradiation at 550 °C. 11
Nuclear Inst. and Methods in Physics Research B 459 (2019) 7–14
Y. Wei, et al.
Fig. 9. Energy spectra of precipitates in specimens: (a) SC-1 irradiated to 15 dpa at 550 °C; (b) SC-2 irradiated to 15 dpa at 550 °C.
Table 1 Compositions (wt%) of precipitates calculated based on EDS spectroscopy for SC-1 (modified with Zr) and SC-2 (modified with Nb, Ta, and W).
S P Si Mn Cr Ni Mo Nb W V Ta Zr Fe
SC-1 Precipitates
SC-2 Precipitates
0.7 0.6 0.8 0.0 20.2 24.9 0.0 – – 0.2 – 0 52.7
5.1 0.0 11.4 0.0 9.8 2.4 1.1 18.4 0.0 0.2 49.9 – 1.4
reference, the difference in recombination rate arising from the binding energy of a vacancy and solute atoms could be as much as two orders of magnitude [29]. It is quite possible that the recombination rate in SC-1 was higher than in SC-2, although the concentration of solute atoms in SC-1 was lower. We thus conclude that at 290 °C, compared with SC-1, a higher release rate and lower recombination rate in SC-2 led to the formation of vacancy clusters with high number density and great sizes, as observed in the present study. According to Eq. (6), the release rate coefficient decreases as the vacancy binding energy of additive atoms increases. It means that additive atoms with smaller binding energy with vacancies act as more effective vacancy cluster nucleation sites and less effective recombination sites. Since vacancy binding energy of Zr is higher than those of three other elements, the ratio of release rate and recombination rate at additive atoms in SC-2 is larger than that in SC-1. The concentration of additive atoms in SC-2 is four times higher than SC-1. It means that capture rate of vacancies by additive atoms in SC-2 is larger than SC-1 according to Eq. (1). It suggests that release rate of vacancy clusters from additive atoms in SC-2 is higher than that in SC-1, which means larger nucleation rate in SC-2. It is consistent to some simulation results in literatures (162.164.168). On the other hand, Zr atoms were more effective recombination sites. It is quite possible that point defect recombination rate in SC-1 is larger than that in SC-2. It means that concentration of vacancies in SC-2 is higher. It is helpful for the formation of vacancy clusters with large sizes and high densities. As the irradiation temperature was increased to 550 °C, SC-1 and SC-2 exhibited different behaviors in response to the temperature change. For SC-1, number density of vacancy clusters with size below about 1.5 nm greatly decreased and vacancy clusters above 2 nm and 3 nm could be observed in specimen irradiated to 15 dpa and 30 dpa respectively. These were not found at 290 °C. Based on Eq. (6), the release coefficient becomes large for a high irradiation temperature. Calculation of 316L with 0.1 at% Zr showed that, as the temperature increased, release rate of vacancies from solution atoms increased, but the recombination rate for vacancies and SIAs at additive sites stayed almost constant [29]. The material used in this calculation had a similar concentration of Zr to that in SC-1 [29]. As is known, recombination processes are dominant below 400 °C in our materials, while release becomes the major mechanism for temperatures above 400 °C [29]. Therefore, in SC-1, the release rate was much higher than recombination rate at 550 °C, and Zr acted mainly as a nucleation site rather than a recombination site. It is also noted that recombination could still be enhanced by Zr atoms, and swelling was thus suppressed when compared with 310S with no additive atoms. DV is also large at high temperatures, which, according to Eq. (4), results in a high capture rate of vacancies by solute atoms. Thus, a large value of capture rate of
paths; besides recombination with SIAs, some vacancies combine with vacancies and vacancy clusters, resulting in nucleation and growth of vacancy clusters, while the remaining vacancies are captured by additive atoms. Eq. (2) implies that vacancy clusters originate from coalescence of migrating vacancies or release of solute atoms. The latter mechanism reflects the enhancement of vacancy cluster nucleation by solute atoms. Eq. (3) suggests that vacancies trapped by additive atoms fall undergo two fates: recombination with interstitial atoms or release from additive atoms as vacancy-clusters. In other words, the solute atoms simultaneously act as both recombination sinks and vacancy cluster nucleation sites. Compared with SC-1, higher number density larger vacancy clusters were observed in SC-2 after 290 °C irradiation. This difference resulted from the variation of composition and concentration of additive atoms. As pointed out in the literature [10], vacancy trapping of oversized solute atoms is due to the size effect of solute atoms, which means that the higher the size factor, the larger is the binding energy of solute atoms with vacancies and vacancy clusters. Based on these literatures [10,30], it is expected that the linear size factors of additive elements rank in the order of lsfZr > lsfTa > lsfNb > lsfW . Here lsfx represents the linear size factor of element x. This implies that, of the four additive elements, Zr has the largest binding energy with vacancies and vacancy clusters [12,29]. Because the concentration of solute atoms in SC-2 was nearly four times higher than in SC-1, hence improving releasing, the release rate in SC-2 was higher than in SC-1, resulting in more effective nucleation of vacancy clusters in SC-2. Some calculation results indicate a small nucleation rate of the clusters for a higher binding energy between additive element and vacancy [18,31]. According to this 12
Nuclear Inst. and Methods in Physics Research B 459 (2019) 7–14
Y. Wei, et al.
vacancies by solute atoms in SC-1 was expected when the temperature increased from 290 °C to 550 °C. High capture rate leads to a low value of CV , which means that nucleation of vacancy clusters is less effective. Calculations in the literature [17] also show that the nucleation rate of vacancy clusters is small at high temperatures for alloys with solute additives, which can explain the observed low number density of small vacancy clusters in SC-1 at 550 °C. Considering the strong interactions between vacancies and Zr atoms, it is quite possible that there are more vacancies in vacancy clusters released from solute atoms. In other words, Zr atoms act not only as nucleation sites for vacancy clusters, but also promote their growth. This accounts for the formation of large vacancy clusters in the specimens. In addition, at a high temperature, the diffusion of vacancy clusters is enhanced. As a result, migration and coalescence of vacancy clusters are more effective, resulting in low number density and large sizes of them. Compared with SC-1, different behavior was observed in SC-2. As Fig. 5 shows, the average size and number density of vacancy clusters in SC-2 both increased slightly after irradiation at 550 °C. As mentioned earlier, Nb and Ta-enriched precipitates were observed in SC-2 irradiated at 550 °C. These caused a reduction in concentration of solute atoms. The decrease of CSol certainly could have reduced the capture rate of vacancies by solute atoms, suppressing nucleation of vacancy clusters and causing a low number density of them. However, recombination of vacancies and SIAs at additive atom sites was also simultaneously suppressed, contributing to a high value of CV . At the same time, a low capture rate also means a high value of CV . Higher temperatures and fewer solute atoms imply higher diffusivity of vacancies and vacancy clusters, leading to larger values of VV and VVn . The combined increases in CV , V Vn C V CVn .
and
Two types of 310S austenitic steels modified by different oversized elements were irradiated with Ar+ ions to damage dose up to 30 dpa at 290 °C and 550 °C respectively. On the bases of the results obtained by SRIM calculations, the kinetic-rate theory models and experimental measurements by TEM technique, the main conclusions are summarized as follows: (1) At 290 °C, Ta, Nb, and W atoms acted as effective nucleation sites for vacancy clusters, but inefficient recombination sites for vacancies and interstitial atoms, giving rise to high number density large vacancy clusters in SC-2; (2) On increasing the temperature from 290 °C to 550 °C, Zr atoms became nucleation sites for vacancy clusters instead of recombination sites for vacancies and SIAs. The capture rate of vacancies by solute atoms increased. At the same time, we believe that Zr atoms enhanced the growth of vacancy clusters because of their strong interaction with vacancies. These mechanisms explain the greatly increased sizes and low number density of vacancy clusters in SC-1 at 550 °C. (3) In SC-2, precipitation of Nb and Ta suppressed the capturing of vacancies from additive atoms and enhanced the diffusivity of vacancies, leading to enhancement of nucleation and growth of vacancy cluster, resulting in slightly higher number density and larger size of them at 550 °C. (4) Precipitates enriched with Nb and Ta were observed in SC-2 after irradiation at 550 °C due to their higher diffusivities than W and Ni. In SC-1, Zr remained dissolved in the matrix even after irradiation at 550 °C, because of its low diffusivity.
resulted in large values of In other words, there was a more effective nuand cleation process and vacancy clusters grew. Not only do these compensate for the consequence of a decreasing nucleation rate, but also contribute to the formation of high number density, large size vacancy clusters, as observed in the specimens. Also, similar to SC-1, migration and coalescence of small vacancy clusters were enhanced by high temperature in SC-2, and large ones formed. In conclusion, the different features of the vacancy clusters in the two materials after 550 °C irradiation mainly resulted from two factors: the enhanced vacancy cluster growth by Zr atoms at high temperature in SC-1 and the precipitation of Nb and Ta in SC-2. As pointed out earlier, the precipitates formed in the two materials after 550 °C irradiation were different. Ni-enriched precipitates were found in SC-1, while Ta and Nb-enriched precipitates were observed in SC-2. Precipitation of W and Zr with concentrations comparable with those of Ta or Nb did not appear in our specimens. As is well known, the growth of precipitates is closely related to the diffusion of solute atoms. It has been reported that diffusivities of Ta and Nb are higher than that of W in body-centered cubic (bcc) Fe at 1050 K [32,33]. In Ref. [34], diffusivity of Zr in bcc-Fe was calculated as being almost the same as that of Fe self-diffusivity at 1050 K [35]. This value is lower than the diffusivities of Ni, Nb and Ta at the same condition. A similar mechanism may also exist in present experimental situation; namely, having higher diffusivity than Zr, Ni precipitated in SC-1; in contrast, in SC-2, the diffusivities of Nb and Ta were both higher than those of W and Ni, leading to Nb- and Ta-enriched precipitates. In present study, Ar ions were employed to create displacement damage. Ar atoms likely remained in the foil as interstitials, which somehow played a role to trap vacancies. This implies that Ar ion irradiation induced diffusion of solute atoms will be less than that induced by self ion (i.e. Fe ions for 310S steels in present study) irradiation or neutron irradiation. In other words, less precipitates might form compared to self ions and neutrons. However, this does not influence the comparison of the precipitation behavior of two modified 310S steels (SC-1 and SC-2), because the irradiation conditions are the same. V 2 V CV
V V ,
5. Summary
V Vn
6. Data availability The raw data required to reproduce these findings cannot be shared at this time as the data also forms part of an ongoing study. The processed data required to reproduce these findings cannot be shared at this time as the data also forms part of an ongoing study. Declaration of Competing Interest The authors declare that they have no known competing financial interests or personal relationships that could have appeared to influence the work reported in this paper. Acknowledgements Financial supports from the National Natural Science Foundation of China (11775162, 11975170 and U1532134) and the International Science & Technology Cooperation Program of China (2015DFR60370) are gratefully acknowledged. The authors would like to thank the Center for Electron Microscopy at Wuhan University, China. We thank Prof. Congxiao Liu, Alabama A&M University, for the technical assistance during the preparation of this manuscript. We acknowledge the assistance of TEM specimen preparation from Center for Nanoscience and Nanotechnology at Wuhan University. Appendix A. Supplementary data Supplementary data to this article can be found online at https:// doi.org/10.1016/j.nimb.2019.08.016. References [1] K.H. Chang, S.M. Chen, T.K. Yeh, J.J. Kai, Effect of dissolved oxygen content on the oxide structure of Alloy 625 in supercritical water environments at 700 °C, Corros. Sci. 81 (2014) 21–26.
13
Nuclear Inst. and Methods in Physics Research B 459 (2019) 7–14
Y. Wei, et al. [2] Y. Behnamian, A. Mostafaei, A. Kohandehghan, B.S. Amirkhiz, D. Serate, W. Zheng, D. Guzonas, M. Chmielus, W. Chen, J.L. Luo, Characterization of oxide scales grown on alloy 310S stainless steel after long term exposure to supercritical water at 500 °C, Mater. Charact. 120 (2016) 273–284. [3] S.J. Zinkle, G.S. Was, Materials challenges in nuclear energy, Acta Mater. 61 (2013) 735–758. [4] D. Guzonas, R. Novotny, Supercritical water-cooled reactor materials – summary of research and open issues, Prog. Nucl. Energy. 77 (2014) 361–372. [5] K.L. Murty, I. Charit, Structural materials for Gen-IV nuclear reactors: challenges and opportunities, J. Nucl. Mater. 383 (2008) 189–195. [6] C. Sun, R. Hui, W. Qu, S. Yick, Progress in corrosion resistant materials for supercritical water reactors, Corros. Sci. 51 (2009) 2508–2523. [7] W. Zheng, D. Guzonas, K.P. Boyle, J. Li, S. Xu, Materials assessment for the Canadian SCWR core concept, JOM 68 (2016) 456–462. [8] F.A. Garner, Recent insights on the swelling and creep of irradiated austenitic alloys, J. Nucl. Mater. 122–123 (1984) 459–471. [9] M.J. Hackett, J.T. Busby, G.S. Was, The mechanism of Zr and Hf in reducing radiation-induced segregation in 316 stainless steel, Metall. Mater. Trans. A: Phys. Metall. Mater. Sci. 39 (2008) 218–224. [10] T. Kato, H. Takahashi, I. Izumiya, Effects of systematic modification with oversized elements on void formation in 316L austenitic stainless steel under electron irradiation, Mater. Trans. JIM 32 (1991) 921–930. [11] F.A. Garner, H.R. Brager, The influence of Mo, Si, P, C, Ti, Cr, Zr and various trace elements on the neutron-induced swelling of AISI 316 stainless steel, J. Nucl. Mater. 155–157 (1988) 833–837. [12] T. Kato, H. Takahashi, M. Izumiya, Grain boundary segregation under electron irradiation in austenitic stainless steels modified with oversized elements, J. Nucl. Mater. 189 (1992) 167–174. [13] J. Gan, E.P. Simonen, S.M. Bruemmer, L. Fournier, B.H. Sencer, G.S. Was, The effect of oversized solute additions on the microstructure of 316SS irradiated with 5 MeV Ni++ions or 3.2 MeV protons, J. Nucl. Mater. 325 (2004) 94–106. [14] Y. Katano, K. Nakata, S. Jitsukawa, T. Aruga, K. Shiraishi, Effects of titanium and zirconium on void swelling in electron-irradiated austenitic stainless steel, J. Nucl. Mater. 133–134 (1985) 530–534. [15] I. Shibahara, N. Akasaka, S. Onose, H. Okada, S. Ukai, Swellling of advanced austenitic stainless steels developed for the environment of heavy neutron exposure, J. Nucl. Mater. 212–215 (1994) 487–491. [16] F.A. Garner, H.R. Brager, R.J. Puigh, Swelling behavior of titanium-modified alloys in EBR-II, J. Nucl. Mater. 133–134 (1985) 535–539. [17] F.A. Garner, W.G. Wolfer, The effect of solute additions on void nucleation, J. Nucl. Mater. 102 (1981) 143–150. [18] L.K. Mansur, M.H. Yoo, The effects of impurity trapping on irradiation-induced swelling and creep, J. Nucl. Mater. 74 (1978) 228–241. [19] L.K. Mansur, Void swelling in metals and alloys under irradiation: an assessment of
the theory, Nucl. Technol. 40 (1978) 5–34. [20] Y. Sekio, S. Yamashita, N. Sakaguchi, H. Takahashi, Effect of additional minor elements on accumulation behavior of point defects under electron irradiation in austenitic stainless steels, Mater. Trans. 55 (2014) 438–442. [21] W.P. Zhang, Z.Y. Shen, R. Tang, S.X. Jin, Y.X. Song, Y.X. Long, Y.X. Wei, X. Zhou, C. Chen, L.P. Guo, Proton-irradiation induced defects in modified 310S steels characterized with positron annihilation spectroscopy and transmission electron microscopy, Nucl. Instrum. Methods Phys. Res. Sect. B 427 (2018) 1–8. [22] D. Wen, B. Jiang, Q. Wang, F. Yu, X. Li, R. Tang, R. Zhang, G. Chen, C. Dong, Influences of Mo/Zr minor-alloying on the phase precipitation behavior in modified 310S austenitic stainless steels at high temperatures, Mater. Des. 128 (2017) 34–46. [23] Z. Zheng, N. Gao, R. Tang, Y. Yu, W. Zhang, Z. Shen, Y. Long, Y. Wei, L. Guo, Evolution of dislocation loops in austenitic stainless steels implanted with high concentration of hydrogen, Philos. Mag. 97 (2017) 2658–2668. [24] J.F. Ziegler, M.D. Ziegler, J.P. Biersack, SRIM – The stopping and range of ions in matter, Nucl. Instrum. Methods Phys. Res. Sect. B: Beam Interact. Mater. Atoms 268 (2010) 1818–1823. [25] ASTME521, Standard Practice for Neutron Radiation Damage Simulation by Charged Particle Irradiation, in: Annu. B. ASTM Stand. Am. Soc. Test. Mater., American Society for Testing and Materials (ASTM), Philadelphia, PA, 1989: p. D-9. [26] R.E. Stoller, M.B. Toloczko, G.S. Was, A.G. Certain, S. Dwaraknath, F.A. Garner, On the use of SRIM for computing radiation damage exposure, Nucl. Instrum. Methods Phys. Res. Sect. B: Beam Interact. Mater. Atoms 310 (2013) 75–80. [27] D.B. Williams, C.B. Carter, Transmission Electron Microscopy, Springer, US, Boston, MA, 2009. [28] T. Jourdan, G. Bencteux, G. Adjanor, Efficient simulation of kinetics of radiation induced defects: a cluster dynamics approach, J. Nucl. Mater. 444 (2014) 298–313. [29] M.J. Hackett, R. Najafabadi, G.S. Was, Modeling solute-vacancy trapping at oversized solutes and its effect on radiation-induced segregation in Fe-Cr-Ni alloys, J. Nucl. Mater. 389 (2009) 279–287. [30] H.W. King, Quantitative size-factors for metallic solid solutions, J. Mater. Sci. 1 (1966) 79–90. [31] L.K. Mansur, Effects of point defect trapping and solute segregation on irradiationinduced swelling and creep, J. Nucl. Mater. 83 (1979) 109–127. [32] S. Huang, D.L. Worthington, M. Asta, V. Ozolins, G. Ghosh, P.K. Liaw, Calculation of impurity diffusivities in α-Fe using first-principles methods, Acta Mater. 58 (2010) 1982–1993. [33] H. Ding, S. Huang, G. Ghosh, P.K. Liaw, M. Asta, A computational study of impurity diffusivities for 5d transition metal solutes in α-Fe, Scr. Mater. 67 (2012) 732–735. [34] D. Murali, B.K. Panigrahi, M.C. Valsakumar, C.S. Sundar, Diffusion of y and Ti/Zr in bcc iron: A first principles study, J. Nucl. Mater. 419 (2011) 208–212. [35] A. Seeger, Lattice vacancies in high-purity α-iron, Phys. Status Solidi Appl. Res. 167 (1998) 289–311.
14