Defect Evolution in H/He Neutral Beam Irradiated W-ZrC Alloy Using Positron Annihilation Spectroscopy
TIAN Xuefen1, LIU Xiang2, GONG Min1, ZHANG Peiyuan1, WANG Kang3, DENG Aihong1()
1.College of Physics, Sichuan University, Chengdu 610064, China 2.Southwestern Institute of Physics, Chengdu 610041, China 3.College of Physics and Electronic Engineering, Heze University, Heze 274015, China
Cite this article:
TIAN Xuefen, LIU Xiang, GONG Min, ZHANG Peiyuan, WANG Kang, DENG Aihong. Defect Evolution in H/He Neutral Beam Irradiated W-ZrC Alloy Using Positron Annihilation Spectroscopy. Acta Metall Sin, 2021, 57(1): 121-128.
Plasma facing materials (PFMs) in future magnetic fusion devices will face various challenges, such as 14.1 MeV neutron and transmutation gas irradiation at high temperatures. W has been considered as one of the most effective candidates for a PFM in recent years. However, pure W exhibits some drawbacks that limit its applications. Conversely, W-ZrC (W-0.5%ZrC, mass fraction) alloy demonstrates excellent performance, such as a relatively low ductile-brittle transition temperature (DBTT), high ductility, and high strength, which will be particularly useful in future fusion reactors. In this work, the Doppler-broadening slow positron beam analysis (DB-SPBA) and SEM were used to characterize the W-ZrC alloy, which had been irradiated by pure H neutral beam or H+6%He (atomic fraction) neutral beam. In the DB-SPBA, parameters S and W were used to characterize the open volume defects in the samples. Under the pure H neutral beam irradiation, the defects were mainly H-V complexes with a large ratio of vacancy to H in the sample at the surface temperature of 850oC. When the surface temperature of the irradiated sample was 1000oC, there was only one kind of vacancy-type defect without any defect damage layer due to the recovery of defect damage in the sample. The surface morphology was smooth and flat at the irradiated sample surface temperature of 1000oC, and the most of pinhole damage structures disappeared compared to the surface temperature of 850oC. The S value in the sample subjected to the H+6%He neutral beam irradiation at the surface temperature of 800oC was larger than that at 700oC because of the increasing vacancy-type defect volume, and defect types were more complex in the 800oC sample. The defect damage layer in the 800oC sample was wider than that in the 700oC sample. Both the 700oC and 800oC samples presented more than one type of defects, but the sample surface damage was significantly more serious at 800oC.
Table 1 Parameters of W-ZrC alloy samples at room temperature
Sample No.
Irradiation condition
Power / (MW·m-2)
Fluence / m-2
Surface temperature / oC
1
H
10
3.4×1024
850
2
H
10
3.4×1024
1000
3
H+6%He
8
6.7×1024
700
4
H+6%He
8
6.7×1024
800
Table 2 Irradiation parameters of W-ZrC alloy samples
Fig.1 SRIM calculation results of H and He implantation profiles in W-ZrC (a), and the corresponding damage profiles (b)
Fig.2 SEM images of nonirradiated sample (a), and samples No.1 (b), No.2 (c), No.3 (d), and No.4 (e)
Fig.3 S-E plots (a), the curve lines corresponding to the best fit with the VEPFIT program of the experimental data；depth profiles of the fitted S parameter (b), and W-S (c) for different samples (S—low momentum annihilation fraction, E—positron energy, W—high momentum annihilation fraction)
Sample No.
S-layer (1)
S-layer (2)
S-layer (3)
1
2.22±0.15
25.5±2.0
50.1±5.6
2
3.78±0.13
-
64.6±4.5
3
4.96±0.45
52.9±6.1
74±17
4
6.2±3.2
48.8±4.3
75±10
Table 3 The effective positron diffusion lengths of W-ZrC alloy samples under different irradiation conditions
1
Tanabe T, Noda N, Nakamura H. Review of high Z materials for PSI applications [J]. J. Nucl. Mater., 1992, 196-198: 11
2
García-Rosales C. Erosion processes in plasma-wall interactions [J]. J. Nucl. Mater., 1994, 211: 202
3
Chuyanov V A. ITER EDA project status [J]. J. Nucl. Mater., 1996, 233-237: 4
4
Causey R, Wilson K, Venhaus T, et al. Tritium retention in tungsten exposed to intense fluxes of 100 eV tritons [J]. J. Nucl. Mater., 1999, 266-269: 467
5
Janeschitz G. Plasma-wall interaction issues in ITER [J]. J. Nucl. Mater. 2001, 290-293: 1
6
Wurster S, Gludovatz B, Pippan R. High temperature fracture experiments on tungsten-rhenium alloys [J]. Int. J. Refract. Met. Hard Mater., 2010, 28: 692
7
Rieth M, Dafferner B. Limitations of W and W-1% La2O3 for use as structural materials [J]. J. Nucl. Mater., 2005, 342: 20
8
Yar M A, Wahlberg S, Bergqvist H, et al. Spark plasma sintering of tungsten-yttrium oxide composites from chemically synthesized nanopowders and microstructural characterization [J]. J. Nucl. Mater., 2011, 412: 227
9
Xie Z M, Miao S, Liu R, et al. Recrystallization and thermal shock fatigue resistance of nanoscale ZrC dispersion strengthened W alloys as plasma-facing components in fusion devices [J]. J. Nucl. Mater., 2017, 496: 41
10
Kurishita H, Amano Y, Kobayashi S, et al. Development of ultra-fine grained W-TiC and their mechanical properties for fusion applications [J]. J. Nucl. Mater., 2007, 367-370: 1453
11
Wang Y K, Miao S, Xie Z M, et al. Thermal stability and mechanical properties of HfC dispersion strengthened W alloys as plasma-facing components in fusion devices [J]. J. Nucl. Mater., 2017, 492: 260
12
Xie Z M, Liu R, Miao S, et al. Extraordinary high ductility/strength of the interface designed bulk W-ZrC alloy plate at relatively low temperature [J]. Sci. Rep., 2015, 5: 16014
13
Xie Z M, Zhang T, Liu R, et al. Grain growth behavior and mechanical properties of zirconium micro-alloyed and nano-size zirconium carbide dispersion strengthened tungsten alloys [J]. Int. J. Refract. Met. Hard Mater., 2015, 51: 180
14
Xie Z M, Liu R, Miao S, et al. High thermal shock resistance of the hot rolled and swaged bulk W-ZrC alloys [J]. J. Nucl. Mater., 2016, 469: 209
15
Ding H L, Xie Z M, Fang Q F, et al. Determination of the DBTT of nanoscale ZrC doped W alloys through amplitude-dependent internal friction technique [J]. Mater. Sci. Eng., 2018, A716: 268
16
Deng H W, Xie Z M, Wang Y K, et al. Mechanical properties and thermal stability of pure W and W-0.5wt%ZrC alloy manufactured with the same technology [J]. Mater. Sci. Eng., 2018, A715: 117
17
Liu R, Xie Z M, Yang J F, et al. Recent progress on the R&D of W-ZrC alloys for plasma facing components in fusion devices [J]. Nucl. Mater. Energy, 2018, 16: 191
18
Liu X, Lian Y, Greuner H, et al. Irradiation effects of hydrogen and helium plasma on different grade tungsten materials [J]. Nucl. Mater. Energy, 2017, 12: 1314
19
Kajita S, Sakaguchi W, Ohno N, et al. Formation process of tungsten nanostructure by the exposure to helium plasma under fusion relevant plasma conditions [J]. Nucl. Fusion, 2009, 49: 095005
20
Baldwin M J, Doerner R P. Formation of helium induced nanostructure ‘fuzz’ on various tungsten grades [J]. J. Nucl. Mater., 2010, 404: 165
21
Baldwin M J, Doerner R P. Helium induced nanoscopic morphology on tungsten under fusion relevant plasma conditions [J]. Nucl. Fusion, 2008, 48: 035001
22
Nishijima D, Ye M Y, Ohno N, et al. Formation mechanism of bubbles and holes on tungsten surface with low-energy and high-flux helium plasma irradiation in NAGDIS-II [J]. J. Nucl. Mater., 2004, 329-323: 1029
23
Nishijima D, Ye M Y, Ohno N, et al. Incident ion energy dependence of bubble formation on tungsten surface with low energy and high flux helium plasma irradiation [J]. J. Nucl. Mater., 2003, 313-316: 97
24
Tokunaga K, Fujiwara T, Ezato K, et al. Effects of high heat flux hydrogen and helium mixture beam irradiation on surface modification and hydrogen retention in tungsten materials [J]. J. Nucl. Mater., 2009, 390-391: 916
25
Miyamoto M, Nishijima D, Baldwin M J, et al. Microscopic damage of tungsten exposed to deuterium-helium mixture plasma in PISCES and its impacts on retention property [J]. J. Nucl. Mater., 2011, 415(): S657
26
Wang S J, Chen Z Q, Wang B, et al. Applied Positron Spectroscopy [M]. Wuhan: Hubei Science and Technology Press, 2008: 39
Yu W Z. Positron Physics and Its Application [M]. Beijing: Science Press, 2002: 441
郁伟中. 正电子物理及其应用 [M]. 北京: 科学出版社, 2002: 441
28
Van Veen A, Schut H, De Vries J, et al. Analysis of positron profiling data by means of “VEPFIT” [J]. AIP Conf. Proc., 1991, 218: 171
29
Greuner H, Maier H, Balden M, et al. Investigation of W components exposed to high thermal and high H/He fluxes [J]. J. Nucl. Mater., 2011, 417: 495
30
Jiang B, Wan F R, Geng W T. Strong hydrogen trapping at helium in tungsten: Density functional theory calculations [J]. Phys. Rev., 2010, 81B: 134112
31
Iwaoka H, Arita M, Horita Z. Hydrogen diffusion in ultrafine-grained palladium: Roles of dislocations and grain boundaries [J]. Acta Mater., 2016, 107: 168
32
Ramachandran R, David C, Magudapathy P, et al. Study of defect complexes and their evolution with temperature in hydrogen and helium irradiated RAFM steel using positron annihilation spectroscopy [J]. Fusion Eng. Des., 2019, 142: 55
33
Myers S M, Besenbacher F, Nørskov J K. Immobilization mechanisms for ion‐implanted deuterium in aluminum [J]. J. Appl. Phys., 1985, 58: 1841
34
Abramov E, Eliezer D. Hydrogen trapping in helium damaged metals: A theoretical approach [J]. J. Mater. Sci., 1992, 27: 2595
