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Effects of Ti element on the microstructural stability of 9Cr–WVTiN reduced activation martensitic steel under ion irradiation
Journal of Nuclear Materials 455 (2014) 37–40
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Journal of Nuclear Materials journal homepage: www.elsevier.com/locate/jnucmat
Effects of Ti element on the microstructural stability of 9Cr–WVTiN reduced activation martensitic steel under ion irradiation Fengfeng Luo a, Liping Guo a,⇑, Shuoxue Jin a, Tiecheng Li a, Jihong Chen a, Jinping Suo b, Feng Yang b, Z. Yao c a Key Laboratory of Artificial Micro- and Nano-Structures of Ministry of Education, Hubei Nuclear Solid Physics Key Laboratory and School of Physics and Technology, Wuhan University, Wuhan 430072, China b State Key Laboratory of Mould Technology, Institute of Materials Science and Engineering, Huazhong University of Science and Technology, Wuhan 430074, China c Department of Mechanical and Materials Engineering, Queen’s University, Kingston K7L 3N6, ON, Canada
a r t i c l e
i n f o
Article history: Available online 20 March 2014
a b s t r a c t Microstructure of 9Cr–WVTiN reduced-activation martensitic steels with two different Ti concentrations irradiated with Fe+, He+ and H+ at 300 °C was studied with transmission electron microscopy. Small dislocation loops were observed in the irradiated steels. The mean size and number density of dislocation loops decreased with the increase of Ti concentration. The segregation of Cr and Fe in carbides was observed in both irradiated steels, and the enrichment of Cr and depletion of Fe were more severe in the low Ti-concentration 9Cr–WVTiN steel. Ó 2014 Elsevier B.V. All rights reserved.
1. Introduction Reduced-activation ferritic/martensitic (RAFM) steels have been considered as main candidate structural materials for the first wall of fusion reactors because of their excellent superior mechanical properties, microstructural stability and swelling resistance [1,2]. Typical RAFM steels are of 9Cr–WVTa type, e.g. F82H [3], EUROFER97 [4], 9Cr2WVTa [5] and CLAM [6], which were produced and tested in Japan, Europe, USA and China respectively. In 9Cr– WVTa RAFM steels, M23C6 (M = Cr, W, Fe) and MX (M = Ta, V; X = C, N) are mentioned as primary precipitates and M23C6 mainly precipitate along prior austenite grain boundaries and packet boundaries [7]. According to the standard molar formation enthalpy of the second phase precipitates, (Ti, V) (C, N) is likely to be a more stable MX phase compared with (Ta, V) C [8]. Thus 9Cr– WVTiN type RAFM steels with dispersed fine MX (M = Ti,V; X = C, N) particles in the matrix have been designed for a more stable structure during irradiation [9]. Ti is a feasible microalloy element in RAFM steels [10]. Addition Ti to RAFM steel might affect its irradiation stability. Lee et al. investigated the effects of Si and Ti on the phase stability and swelling of AISI 316 steel [11]. Shibayama et al. studied the effect of Ti element on the precipitation of Fe–9Cr–2W steels [12]. However the effect of Ti element on structure stability of 9Cr–WVTiN under irradiation was not clear.
⇑ Corresponding author. Tel.: +86 27 6875 2481x2223; fax: +86 27 6875 2569. E-mail address: [email protected] (L. Guo). http://dx.doi.org/10.1016/j.jnucmat.2014.03.020 0022-3115/Ó 2014 Elsevier B.V. All rights reserved.
In fusion conditions, 14 MeV neutron irradiation induces severe displacement damage and meanwhile produces transmuted helium and hydrogen at a high production rate [13]. As large damage dose, any desired He/dpa and H/dpa ratio could be easily achieved under accurately controlled temperature and other conditions by ion irradiation, self-ion, helium and hydrogen ion irradiation have been used to simulate neutron irradiation effects. In present study, two 9Cr–WVTiN steels with different Ti concentration are irradiated with Fe, He and H ions at 300 °C. Microstructures and compositions of these samples were investigated by Transmission electron microscopy (TEM) and energy dispersive X-ray (EDX), respectively. The objective of this paper is to study the effect of Ti element on the structural stability under irradiation. 2. Experimental The chemical composition of the two tested 9Cr–WVTiN reduced activation martensitic steels (named as SCRAM steels) supplied by Huazhong University of Science and Technology are listed in Table 1 [9]. Both steels had twice quenching and tempering processes, including quenching after annealing at 980 °C for 0.5 h and tempering at 760 °C for 2 h, and then quenching after annealing at 960 °C for 0.5 h and tempering at 740 °C for 2 h. The bulk materials were first cut into 0.5 mm thick sheets and then thinned to about 0.1 mm thick by mechanically polishing. Standard TEM discs in 3 mm diameter were punched out from foils and milled to final thickness 50–100 lm with silicon carbide paper. Finally, TEM specimens were polished by a MTPA-5 twin-jet electropolishing machine (produced by Shanghai Jiaotong University,
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F. Luo et al. / Journal of Nuclear Materials 455 (2014) 37–40
Table 1 Chemical compositions of two RAFM steels in wt%. Material
C
Si
Mn
Cr
W
V
Ti
P
S
O
N
Fe
SCRAM-01 SCRAM-03
0.096 0.095
0.19 0.3
0.45 0.51
9.27 9.21
2.14 2.28
0.095 0.092
0.006 0.018
0.005 0.005
0.005 0.001
0.0041 0.0052
0.0077 0.011
Bal. Bal.
China), using 5% perchloric acid and 95% ethanol polishing solution at 30 °C. The two steels were irradiated under the same condition, using an ion implanter in the Accelerator Laboratory of Wuhan University [14]. Irradiations were performed with three different kinds of ions, firstly using 100 keV Fe+ to a fluence of 1 1016 ions/ cm2, followed by 100 keV He+ to a fluence of 5 1016 ions/cm2, and finally 50 keV H+ to a fluence of 1 1017 ions/cm2. The specimen temperature was maintained at 300 ± 5 °C, which was monitored by a thermocouple during irradiation. The probe of the thermocouple touched the ion incident plane of specimen which was mounted on top of a heater by using fulmargin. The investigation on the microstructure of the steels was performed with a Philips CM20 Transmission Electron Microscope operated at 200 kV. A JEM-2010FEF field-emission-gun analytical TEM equipped with an energy dispersive X-ray (EDX) analyzer were used to measure the compositions of the specimens. 3. Results and discussion 3.1. Unirradiated condition The microstructure of unirradiated SCRAM-01 and SCRAM-03 steels is shown in Fig. 1, in which we can observe martensitic laths and various precipitate particles. The investigated precipitates had been identified to be M23C6 by using electron diffraction. The mean size of precipitates was analyzed by Image-Pro Plus software, using TEM images treated with Photoshop software. Three TEM images from different field of views were used to determine the lath width of martensites in the steels. The detailed parameters of unirradiated steels are summarized in Table 2. The average diameter of precipitates is 105.2 nm for SCRAM-01 steel and 98.1 nm for SCRAM-03 steel, and the corresponding average lath width is 0.62 lm and 0.58 lm, respectively. Compared with SCRAM-01 steel, the mean size of precipitates and lath width of martensites in SCRAM-03 steel which contains more Ti element is a bit smaller.
Table 2 A summary of microstructure analysis for samples before and after ion irradiation. Sample
Precipitates (nm) Lath width (lm)
SCRAM-01 steel
SCRAM-03 steel
Unirradiated
Irradiated
Unirradiated
Irradiated
105.2 ± 3.6 0.62 ± 0.03
106.6 ± 4.2 0.81 ± 0.04
98.1 ± 5.1 0.58 ± 0.03
99.9 ± 4.3 0.68 ± 0.02
3.2. Dislocation loops Weak-beam dark-field (WEDF) micrographs of the irradiated SCRAM-01 and SCRAM-03 steels are shown in Fig. 2, in which small dislocation loops are observed. The thickness of observation area of specimen was fixed at 30–50 nm, which was estimated by counting the number of thickness fringes from the edge of the thin foil. In both case, WEDF micrographs were taken using g = 011 and 0 0 2 in a foil oriented near (1 0 0). Fig. 3 shows the size distribution and density of dislocation loops, which is analyzed in the same WEDF condition. The mean diameter of dislocation loops of irradiated SCARM-01 steel and SCARM-03 steel is 2.9 nm and 2.8 nm, respectively. The number density of dislocation loops was calculated from WEDF micrographs using a pair g = 011 and 0 0 2 in a foil orientation near pole (1 0 0) as follows [15]: qh100i = 2qg=011 qg=002, q1/2h111i = 2/3(qg=011 + 2qg=002). The total number density of dislocation loops including h1 0 0i and 1/ 2h1 1 1i types is 3.8 1017/cm3 for irradiated SCRAM-01 steel and 3.0 1017/cm3 for irradiated SCRAM-03 steel. While the mean loop size of SCRAM-03 is slightly smaller, the number density of loops in SCRAM-03 is less than that of SCRAM-01. The results mean that the addition of more Ti could reduce the formation of irradiationinduced dislocation loops. It is well established that dislocation loops are assumed to act as barriers to gliding dislocations in the slip plane and impede their movements, therefore resulting in radiation hardening [16]. Radiation hardening due to the dislocation loops could be calculated using the dispersed barrier-hardening (DBH) mode [17,18]: Dr = Malb(Nd)0.5, where M is the Taylor factor, a is the barrier strength factor, l is the shear modulus, b is the Burgers vector, N is the number density of dislocation loops and d is the average size for the loops. The hardness changes were calculated using the follow parameters: M = 3.06 for bcc metals [19], a = 0.3 for loops and other defect clusters [20], l = 8.0 101 GPa, b = 2.68 101 nm [21]. The equation showed that the yield strength change is proportional to the square root of the product of the defect density and size. According to the measured number density and average diameter of dislocation loops, the increment of hardness is 653 MPa for SCRAM-01 steel and 570 MPa for SCRAM-03 steel. Compared with the low Ti-concentration SCRAM-01 steel, the increment of hardness is estimated to be smaller in the high Ti-concentration steel SCRAM-03. 3.3. Precipitates
Fig. 1. Microstructure of unirradiated (a) SCRAM-01 steel and (b) SCRAM-03 steel.
Typical bright-field micrograph of SCRAM-01 and SCRAM-03 steels after ion irradiation is shown in Fig. 4. The detailed parameters of irradiated samples are also summarized in Table 2. The
F. Luo et al. / Journal of Nuclear Materials 455 (2014) 37–40
39
Fig. 2. Dislocation structures in irradiated (a and b) SCRAM-01 steel and (c and d) SCRAM-03 steel using the weak beam condition: s > 0, g = 011 and 0 0 2 near pole [1 0 0].
Fig. 3. Dislocation loop distribution of irradiated SCRAM-01 and SCRAM-03 steels analyzed under the same weak beam condition: s > 0, g = 110 (g, 3g), near pole [1 1 1].
average size of precipitates and the width of martensitic lath become larger in both steels after irradiation. The average size of precipitates changes little (<2%) in two steels after irradiation. However, the difference between the two steels is manifest: the increment of lath width after irradiation is 0.19 lm (31%) for SCRAM-01 steel but 0.10 lm (15%) for SCRAM-03 steel. Fig. 5 shows EDX maps for iron, chromium, manganese and tungsten obtained by using a scanning transmission electron microscope (STEM) in the region contained carbide precipitates for irradiated SCARM-01 and SCRAM-03 steels. For unirradiated and irradiated specimens of SCRAM-01 steel, 18 M23C6 precipitates were identified and analyzed; for unirradiated and irradiated specimens of SCRAM-03 steel, 15 M23C6 precipitates were identified and analyzed. Detailed compositions of carbides and matrices in unirradiated and irradiated steels are shown in Table 3. The concentration of Fe and Cr changed obviously after irradiation in carbide precipitates, while the composition of matrices changed a little in both steels. These results show that apparent enrichment
Fig. 4. Microstructure of (a) SCRAM-01 steel and (b) SCRAM-03 steel irradiated at 300 °C.
of chromium element and tungsten element and depletion of iron element occurred in carbide precipitates. Lu et al. gave a possible mechanism for Cr segregation behavior [22]: Cr atoms in ferritic/ martensitic steels may be oversized or undersized. Since the steels used here are martensitic steels, the Cr atoms may change from oversized to undersized. For undersized Cr atoms, Cr-self-interstitial Fe complexes may be preferentially formed by the positive binding energy and Cr atoms will be solute-dragged towards carbides and therefore result in the enrichment of Cr in carbides. However this mechanism is not suitable for the enrichment of W in carbides and it requires further investigation. The segregation of elements could influence mechanical properties of RAFM steels. The enrichment of Cr could affect the creep strength of steel and induce additional precipitate phase and the enrichment of W would reduce the high temperature strength [23]. From Table 3, we could know that the enrichment of Cr and the depletion of Fe were little more strongly in carbides of SCRAM-01 steel than of SCRAM-03 steel, indicating that SCRAM03 steel is likely to be more stable than SCRAM-01 steel during irradiation.
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F. Luo et al. / Journal of Nuclear Materials 455 (2014) 37–40
Fig. 5. STEM investigation and EDX maps for each element of irradiated (a) SCRAM-01 steel and (b) SCRAM-03 steel.
Table 3 EDX results of unirradiated and irradiated specimens (wt%). Elements
SCRAM-01 steel
SCRAM-03 steel
Unirradiated
Fe Cr W Mn
Irradiated
Unirradiated
Irradiated
M
C
CH
M
C
CH
M
C
CH
M
C
CH
86.29 8.10 4.93 0.68
67.81 22.08 10.11 0.00
18.48 13.98 5.18 0.68
87.38 9.26 2.59 0.77
46.98 38.80 13.70 0.52
40.40 29.54 11.11 0.25
86.74 7.77 4.97 0.52
72.25 17.84 9.86 0.05
14.49 10.07 4.89 0.47
87.22 9.28 2.80 0.70
52.49 34.74 12.02 0.75
34.73 25.46 9.22 0.05
M refers to Matrices; C refers to Carbides; CH refers to the change in carbides composition from matrices composition. The bold values refer to the change in carbides composition from matrices composition.
4. Conclusions
References
Microstructural changes of two 9Cr–WVTiN steels with different Ti concentration under ion radiations at 300 °C were investigated in present study. The main conclusions are drawn as follows:
[1] N. Baluc, R. Schäublin, P. Spätig, M. Victoria, Nucl. Fusion 44 (2004) 56. [2] T. Muroga, M. Gasparotto, S. Zinkle, Fusion Eng. Des. 61 (2002) 13. [3] S. Jitsukawa, M. Tamura, B. Van der Schaaf, R. Klueh, A. Alamo, C. Petersen, M. Schirra, P. Spaetig, G. Odette, A. Tavassoli, J. Nucl. Mater. 307 (2002) 179. [4] R. Lindau, A. Möslang, M. Rieth, M. Klimiankou, E. Materna-Morris, A. Alamo, A.-A. Tavassoli, C. Cayron, A.-M. Lancha, P. Fernandez, Fusion Eng. Des. 75 (2005) 989. [5] R. Klueh, D. Alexander, E. Kenik, J. Nucl. Mater. 227 (1995) 11. [6] Q. Huang, C. Li, Y. Li, M. Chen, M. Zhang, L. Peng, Z. Zhu, Y. Song, S. Gao, J. Nucl. Mater. 367 (2007) 142. [7] P. Ennis, A. Zielinska-Lipiec, O. Wachter, A. Czyrska-Filemonowicz, Acta Mater. 45 (1997) 4901. [8] Q.L. Yong, Secondary Phases in Steels, first ed., Metallurgical Industry Press, Beijing, 2006. [9] X. Xiong, F. Yang, X. Zou, J. Suo, J. Nucl. Mater. 430 (2012) 114. [10] A. Kohyama, A. Hishinuma, D. Gelles, R. Klueh, W. Dietz, K. Ehrlich, J. Nucl. Mater. 233 (1996) 138. [11] E. Lee, A. Rowcliffe, E. Kenik, J. Nucl. Mater. 83 (1979) 79. [12] T. Shibayama, A. Kimura, H. Kayano, J. Nucl. Mater. 233 (1996) 270. [13] M.S. Wechsler, D.R. Davidson, L.R. Greenwood, W.F. Sommer, Effects of Radiation on Materials: 12th Conference, ASTM, 1985. [14] L. Guo, C. Liu, M. Li, B. Song, M. Ye, D. Fu, X. Fan, Nucl. Instrum. Methods Phys. Res., A 586 (2008) 143. [15] H. Ogiwara, A. Kohyama, H. Tanigawa, H. Sakasegawa, Fusion Eng. Des. 81 (2006) 1091. [16] C. Pokor, X. Averty, Y. Brechet, P. Dubuisson, J. Massoud, Scr. Mater. 50 (2004) 597. [17] A. Seeger, in: Proceedings of the Second UN International Conference on the Peaceful Uses of Atomic Energy, vol. 6, New York United Nations, 1958, p. 250. [18] B. Singh, A. Foreman, H. Trinkaus, J. Nucl. Mater. 249 (1997) 103. [19] S.J. Zinkle, Y. Matsukawa, J. Nucl. Mater. 329 (2004) 88. [20] E. Wakai, A. Hishinuma, K. Usami, Y. Kato, S. Takaki, K. Abiko, Mater. Trans. 41 (2000) 1180. [21] R. Schaeublin, D. Gelles, M. Victoria, J. Nucl. Mater. 307 (2002) 197. [22] Z. Lu, R. Faulkner, G. Was, B. Wirth, Scr. Mater. 58 (2008) 878. [23] S. Jin, L. Guo, Z. Yang, D. Fu, C. Liu, R. Tang, F. Liu, Y. Qiao, H. Zhang, Mater. Charact. 62 (2011) 136.
(1) Small dislocation loops were observed in both irradiated steels. The mean size and number density of the loops decreased with the increasing Ti concentration. The hardness increment due to the formation of dislocation loops of the low Ti-concentration 9Cr–WVTiN steel were estimated to be larger than high Ti-concentration 9Cr–WVTiN steel by the dispersed barrier-hardening model. (2) After irradiation, the average size of precipitates and width of martensitic laths become larger in both steels, the size of precipitates changes little, however, the increment of lath width in low Ti-concentration steel is obviously larger. (3) The segregation of Cr and Fe in carbides was observed in both irradiated steels, and the enrichment of Cr and depletion of Fe were more severe in low Ti-concentration 9Cr– WVTiN steel.
Acknowledgements The financial supports from the National Magnetic Confinement Fusion Program (2011GB108009), the National Natural Science Foundation of China (11075119 and 11275140) and the Fundamental Research Funds for the Central Universities (2012202020207) are gratefully acknowledged. National Science and Engineering Research Council of Canada (NSERC) is acknowledged for supporting authors using facility in Queen’s University.
Contents lists available at ScienceDirect
Journal of Nuclear Materials journal homepage: www.elsevier.com/locate/jnucmat
Effects of Ti element on the microstructural stability of 9Cr–WVTiN reduced activation martensitic steel under ion irradiation Fengfeng Luo a, Liping Guo a,⇑, Shuoxue Jin a, Tiecheng Li a, Jihong Chen a, Jinping Suo b, Feng Yang b, Z. Yao c a Key Laboratory of Artificial Micro- and Nano-Structures of Ministry of Education, Hubei Nuclear Solid Physics Key Laboratory and School of Physics and Technology, Wuhan University, Wuhan 430072, China b State Key Laboratory of Mould Technology, Institute of Materials Science and Engineering, Huazhong University of Science and Technology, Wuhan 430074, China c Department of Mechanical and Materials Engineering, Queen’s University, Kingston K7L 3N6, ON, Canada
a r t i c l e
i n f o
Article history: Available online 20 March 2014
a b s t r a c t Microstructure of 9Cr–WVTiN reduced-activation martensitic steels with two different Ti concentrations irradiated with Fe+, He+ and H+ at 300 °C was studied with transmission electron microscopy. Small dislocation loops were observed in the irradiated steels. The mean size and number density of dislocation loops decreased with the increase of Ti concentration. The segregation of Cr and Fe in carbides was observed in both irradiated steels, and the enrichment of Cr and depletion of Fe were more severe in the low Ti-concentration 9Cr–WVTiN steel. Ó 2014 Elsevier B.V. All rights reserved.
1. Introduction Reduced-activation ferritic/martensitic (RAFM) steels have been considered as main candidate structural materials for the first wall of fusion reactors because of their excellent superior mechanical properties, microstructural stability and swelling resistance [1,2]. Typical RAFM steels are of 9Cr–WVTa type, e.g. F82H [3], EUROFER97 [4], 9Cr2WVTa [5] and CLAM [6], which were produced and tested in Japan, Europe, USA and China respectively. In 9Cr– WVTa RAFM steels, M23C6 (M = Cr, W, Fe) and MX (M = Ta, V; X = C, N) are mentioned as primary precipitates and M23C6 mainly precipitate along prior austenite grain boundaries and packet boundaries [7]. According to the standard molar formation enthalpy of the second phase precipitates, (Ti, V) (C, N) is likely to be a more stable MX phase compared with (Ta, V) C [8]. Thus 9Cr– WVTiN type RAFM steels with dispersed fine MX (M = Ti,V; X = C, N) particles in the matrix have been designed for a more stable structure during irradiation [9]. Ti is a feasible microalloy element in RAFM steels [10]. Addition Ti to RAFM steel might affect its irradiation stability. Lee et al. investigated the effects of Si and Ti on the phase stability and swelling of AISI 316 steel [11]. Shibayama et al. studied the effect of Ti element on the precipitation of Fe–9Cr–2W steels [12]. However the effect of Ti element on structure stability of 9Cr–WVTiN under irradiation was not clear.
⇑ Corresponding author. Tel.: +86 27 6875 2481x2223; fax: +86 27 6875 2569. E-mail address: [email protected] (L. Guo). http://dx.doi.org/10.1016/j.jnucmat.2014.03.020 0022-3115/Ó 2014 Elsevier B.V. All rights reserved.
In fusion conditions, 14 MeV neutron irradiation induces severe displacement damage and meanwhile produces transmuted helium and hydrogen at a high production rate [13]. As large damage dose, any desired He/dpa and H/dpa ratio could be easily achieved under accurately controlled temperature and other conditions by ion irradiation, self-ion, helium and hydrogen ion irradiation have been used to simulate neutron irradiation effects. In present study, two 9Cr–WVTiN steels with different Ti concentration are irradiated with Fe, He and H ions at 300 °C. Microstructures and compositions of these samples were investigated by Transmission electron microscopy (TEM) and energy dispersive X-ray (EDX), respectively. The objective of this paper is to study the effect of Ti element on the structural stability under irradiation. 2. Experimental The chemical composition of the two tested 9Cr–WVTiN reduced activation martensitic steels (named as SCRAM steels) supplied by Huazhong University of Science and Technology are listed in Table 1 [9]. Both steels had twice quenching and tempering processes, including quenching after annealing at 980 °C for 0.5 h and tempering at 760 °C for 2 h, and then quenching after annealing at 960 °C for 0.5 h and tempering at 740 °C for 2 h. The bulk materials were first cut into 0.5 mm thick sheets and then thinned to about 0.1 mm thick by mechanically polishing. Standard TEM discs in 3 mm diameter were punched out from foils and milled to final thickness 50–100 lm with silicon carbide paper. Finally, TEM specimens were polished by a MTPA-5 twin-jet electropolishing machine (produced by Shanghai Jiaotong University,
38
F. Luo et al. / Journal of Nuclear Materials 455 (2014) 37–40
Table 1 Chemical compositions of two RAFM steels in wt%. Material
C
Si
Mn
Cr
W
V
Ti
P
S
O
N
Fe
SCRAM-01 SCRAM-03
0.096 0.095
0.19 0.3
0.45 0.51
9.27 9.21
2.14 2.28
0.095 0.092
0.006 0.018
0.005 0.005
0.005 0.001
0.0041 0.0052
0.0077 0.011
Bal. Bal.
China), using 5% perchloric acid and 95% ethanol polishing solution at 30 °C. The two steels were irradiated under the same condition, using an ion implanter in the Accelerator Laboratory of Wuhan University [14]. Irradiations were performed with three different kinds of ions, firstly using 100 keV Fe+ to a fluence of 1 1016 ions/ cm2, followed by 100 keV He+ to a fluence of 5 1016 ions/cm2, and finally 50 keV H+ to a fluence of 1 1017 ions/cm2. The specimen temperature was maintained at 300 ± 5 °C, which was monitored by a thermocouple during irradiation. The probe of the thermocouple touched the ion incident plane of specimen which was mounted on top of a heater by using fulmargin. The investigation on the microstructure of the steels was performed with a Philips CM20 Transmission Electron Microscope operated at 200 kV. A JEM-2010FEF field-emission-gun analytical TEM equipped with an energy dispersive X-ray (EDX) analyzer were used to measure the compositions of the specimens. 3. Results and discussion 3.1. Unirradiated condition The microstructure of unirradiated SCRAM-01 and SCRAM-03 steels is shown in Fig. 1, in which we can observe martensitic laths and various precipitate particles. The investigated precipitates had been identified to be M23C6 by using electron diffraction. The mean size of precipitates was analyzed by Image-Pro Plus software, using TEM images treated with Photoshop software. Three TEM images from different field of views were used to determine the lath width of martensites in the steels. The detailed parameters of unirradiated steels are summarized in Table 2. The average diameter of precipitates is 105.2 nm for SCRAM-01 steel and 98.1 nm for SCRAM-03 steel, and the corresponding average lath width is 0.62 lm and 0.58 lm, respectively. Compared with SCRAM-01 steel, the mean size of precipitates and lath width of martensites in SCRAM-03 steel which contains more Ti element is a bit smaller.
Table 2 A summary of microstructure analysis for samples before and after ion irradiation. Sample
Precipitates (nm) Lath width (lm)
SCRAM-01 steel
SCRAM-03 steel
Unirradiated
Irradiated
Unirradiated
Irradiated
105.2 ± 3.6 0.62 ± 0.03
106.6 ± 4.2 0.81 ± 0.04
98.1 ± 5.1 0.58 ± 0.03
99.9 ± 4.3 0.68 ± 0.02
3.2. Dislocation loops Weak-beam dark-field (WEDF) micrographs of the irradiated SCRAM-01 and SCRAM-03 steels are shown in Fig. 2, in which small dislocation loops are observed. The thickness of observation area of specimen was fixed at 30–50 nm, which was estimated by counting the number of thickness fringes from the edge of the thin foil. In both case, WEDF micrographs were taken using g = 011 and 0 0 2 in a foil oriented near (1 0 0). Fig. 3 shows the size distribution and density of dislocation loops, which is analyzed in the same WEDF condition. The mean diameter of dislocation loops of irradiated SCARM-01 steel and SCARM-03 steel is 2.9 nm and 2.8 nm, respectively. The number density of dislocation loops was calculated from WEDF micrographs using a pair g = 011 and 0 0 2 in a foil orientation near pole (1 0 0) as follows [15]: qh100i = 2qg=011 qg=002, q1/2h111i = 2/3(qg=011 + 2qg=002). The total number density of dislocation loops including h1 0 0i and 1/ 2h1 1 1i types is 3.8 1017/cm3 for irradiated SCRAM-01 steel and 3.0 1017/cm3 for irradiated SCRAM-03 steel. While the mean loop size of SCRAM-03 is slightly smaller, the number density of loops in SCRAM-03 is less than that of SCRAM-01. The results mean that the addition of more Ti could reduce the formation of irradiationinduced dislocation loops. It is well established that dislocation loops are assumed to act as barriers to gliding dislocations in the slip plane and impede their movements, therefore resulting in radiation hardening [16]. Radiation hardening due to the dislocation loops could be calculated using the dispersed barrier-hardening (DBH) mode [17,18]: Dr = Malb(Nd)0.5, where M is the Taylor factor, a is the barrier strength factor, l is the shear modulus, b is the Burgers vector, N is the number density of dislocation loops and d is the average size for the loops. The hardness changes were calculated using the follow parameters: M = 3.06 for bcc metals [19], a = 0.3 for loops and other defect clusters [20], l = 8.0 101 GPa, b = 2.68 101 nm [21]. The equation showed that the yield strength change is proportional to the square root of the product of the defect density and size. According to the measured number density and average diameter of dislocation loops, the increment of hardness is 653 MPa for SCRAM-01 steel and 570 MPa for SCRAM-03 steel. Compared with the low Ti-concentration SCRAM-01 steel, the increment of hardness is estimated to be smaller in the high Ti-concentration steel SCRAM-03. 3.3. Precipitates
Fig. 1. Microstructure of unirradiated (a) SCRAM-01 steel and (b) SCRAM-03 steel.
Typical bright-field micrograph of SCRAM-01 and SCRAM-03 steels after ion irradiation is shown in Fig. 4. The detailed parameters of irradiated samples are also summarized in Table 2. The
F. Luo et al. / Journal of Nuclear Materials 455 (2014) 37–40
39
Fig. 2. Dislocation structures in irradiated (a and b) SCRAM-01 steel and (c and d) SCRAM-03 steel using the weak beam condition: s > 0, g = 011 and 0 0 2 near pole [1 0 0].
Fig. 3. Dislocation loop distribution of irradiated SCRAM-01 and SCRAM-03 steels analyzed under the same weak beam condition: s > 0, g = 110 (g, 3g), near pole [1 1 1].
average size of precipitates and the width of martensitic lath become larger in both steels after irradiation. The average size of precipitates changes little (<2%) in two steels after irradiation. However, the difference between the two steels is manifest: the increment of lath width after irradiation is 0.19 lm (31%) for SCRAM-01 steel but 0.10 lm (15%) for SCRAM-03 steel. Fig. 5 shows EDX maps for iron, chromium, manganese and tungsten obtained by using a scanning transmission electron microscope (STEM) in the region contained carbide precipitates for irradiated SCARM-01 and SCRAM-03 steels. For unirradiated and irradiated specimens of SCRAM-01 steel, 18 M23C6 precipitates were identified and analyzed; for unirradiated and irradiated specimens of SCRAM-03 steel, 15 M23C6 precipitates were identified and analyzed. Detailed compositions of carbides and matrices in unirradiated and irradiated steels are shown in Table 3. The concentration of Fe and Cr changed obviously after irradiation in carbide precipitates, while the composition of matrices changed a little in both steels. These results show that apparent enrichment
Fig. 4. Microstructure of (a) SCRAM-01 steel and (b) SCRAM-03 steel irradiated at 300 °C.
of chromium element and tungsten element and depletion of iron element occurred in carbide precipitates. Lu et al. gave a possible mechanism for Cr segregation behavior [22]: Cr atoms in ferritic/ martensitic steels may be oversized or undersized. Since the steels used here are martensitic steels, the Cr atoms may change from oversized to undersized. For undersized Cr atoms, Cr-self-interstitial Fe complexes may be preferentially formed by the positive binding energy and Cr atoms will be solute-dragged towards carbides and therefore result in the enrichment of Cr in carbides. However this mechanism is not suitable for the enrichment of W in carbides and it requires further investigation. The segregation of elements could influence mechanical properties of RAFM steels. The enrichment of Cr could affect the creep strength of steel and induce additional precipitate phase and the enrichment of W would reduce the high temperature strength [23]. From Table 3, we could know that the enrichment of Cr and the depletion of Fe were little more strongly in carbides of SCRAM-01 steel than of SCRAM-03 steel, indicating that SCRAM03 steel is likely to be more stable than SCRAM-01 steel during irradiation.
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F. Luo et al. / Journal of Nuclear Materials 455 (2014) 37–40
Fig. 5. STEM investigation and EDX maps for each element of irradiated (a) SCRAM-01 steel and (b) SCRAM-03 steel.
Table 3 EDX results of unirradiated and irradiated specimens (wt%). Elements
SCRAM-01 steel
SCRAM-03 steel
Unirradiated
Fe Cr W Mn
Irradiated
Unirradiated
Irradiated
M
C
CH
M
C
CH
M
C
CH
M
C
CH
86.29 8.10 4.93 0.68
67.81 22.08 10.11 0.00
18.48 13.98 5.18 0.68
87.38 9.26 2.59 0.77
46.98 38.80 13.70 0.52
40.40 29.54 11.11 0.25
86.74 7.77 4.97 0.52
72.25 17.84 9.86 0.05
14.49 10.07 4.89 0.47
87.22 9.28 2.80 0.70
52.49 34.74 12.02 0.75
34.73 25.46 9.22 0.05
M refers to Matrices; C refers to Carbides; CH refers to the change in carbides composition from matrices composition. The bold values refer to the change in carbides composition from matrices composition.
4. Conclusions
References
Microstructural changes of two 9Cr–WVTiN steels with different Ti concentration under ion radiations at 300 °C were investigated in present study. The main conclusions are drawn as follows:
[1] N. Baluc, R. Schäublin, P. Spätig, M. Victoria, Nucl. Fusion 44 (2004) 56. [2] T. Muroga, M. Gasparotto, S. Zinkle, Fusion Eng. Des. 61 (2002) 13. [3] S. Jitsukawa, M. Tamura, B. Van der Schaaf, R. Klueh, A. Alamo, C. Petersen, M. Schirra, P. Spaetig, G. Odette, A. Tavassoli, J. Nucl. Mater. 307 (2002) 179. [4] R. Lindau, A. Möslang, M. Rieth, M. Klimiankou, E. Materna-Morris, A. Alamo, A.-A. Tavassoli, C. Cayron, A.-M. Lancha, P. Fernandez, Fusion Eng. Des. 75 (2005) 989. [5] R. Klueh, D. Alexander, E. Kenik, J. Nucl. Mater. 227 (1995) 11. [6] Q. Huang, C. Li, Y. Li, M. Chen, M. Zhang, L. Peng, Z. Zhu, Y. Song, S. Gao, J. Nucl. Mater. 367 (2007) 142. [7] P. Ennis, A. Zielinska-Lipiec, O. Wachter, A. Czyrska-Filemonowicz, Acta Mater. 45 (1997) 4901. [8] Q.L. Yong, Secondary Phases in Steels, first ed., Metallurgical Industry Press, Beijing, 2006. [9] X. Xiong, F. Yang, X. Zou, J. Suo, J. Nucl. Mater. 430 (2012) 114. [10] A. Kohyama, A. Hishinuma, D. Gelles, R. Klueh, W. Dietz, K. Ehrlich, J. Nucl. Mater. 233 (1996) 138. [11] E. Lee, A. Rowcliffe, E. Kenik, J. Nucl. Mater. 83 (1979) 79. [12] T. Shibayama, A. Kimura, H. Kayano, J. Nucl. Mater. 233 (1996) 270. [13] M.S. Wechsler, D.R. Davidson, L.R. Greenwood, W.F. Sommer, Effects of Radiation on Materials: 12th Conference, ASTM, 1985. [14] L. Guo, C. Liu, M. Li, B. Song, M. Ye, D. Fu, X. Fan, Nucl. Instrum. Methods Phys. Res., A 586 (2008) 143. [15] H. Ogiwara, A. Kohyama, H. Tanigawa, H. Sakasegawa, Fusion Eng. Des. 81 (2006) 1091. [16] C. Pokor, X. Averty, Y. Brechet, P. Dubuisson, J. Massoud, Scr. Mater. 50 (2004) 597. [17] A. Seeger, in: Proceedings of the Second UN International Conference on the Peaceful Uses of Atomic Energy, vol. 6, New York United Nations, 1958, p. 250. [18] B. Singh, A. Foreman, H. Trinkaus, J. Nucl. Mater. 249 (1997) 103. [19] S.J. Zinkle, Y. Matsukawa, J. Nucl. Mater. 329 (2004) 88. [20] E. Wakai, A. Hishinuma, K. Usami, Y. Kato, S. Takaki, K. Abiko, Mater. Trans. 41 (2000) 1180. [21] R. Schaeublin, D. Gelles, M. Victoria, J. Nucl. Mater. 307 (2002) 197. [22] Z. Lu, R. Faulkner, G. Was, B. Wirth, Scr. Mater. 58 (2008) 878. [23] S. Jin, L. Guo, Z. Yang, D. Fu, C. Liu, R. Tang, F. Liu, Y. Qiao, H. Zhang, Mater. Charact. 62 (2011) 136.
(1) Small dislocation loops were observed in both irradiated steels. The mean size and number density of the loops decreased with the increasing Ti concentration. The hardness increment due to the formation of dislocation loops of the low Ti-concentration 9Cr–WVTiN steel were estimated to be larger than high Ti-concentration 9Cr–WVTiN steel by the dispersed barrier-hardening model. (2) After irradiation, the average size of precipitates and width of martensitic laths become larger in both steels, the size of precipitates changes little, however, the increment of lath width in low Ti-concentration steel is obviously larger. (3) The segregation of Cr and Fe in carbides was observed in both irradiated steels, and the enrichment of Cr and depletion of Fe were more severe in low Ti-concentration 9Cr– WVTiN steel.
Acknowledgements The financial supports from the National Magnetic Confinement Fusion Program (2011GB108009), the National Natural Science Foundation of China (11075119 and 11275140) and the Fundamental Research Funds for the Central Universities (2012202020207) are gratefully acknowledged. National Science and Engineering Research Council of Canada (NSERC) is acknowledged for supporting authors using facility in Queen’s University.