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Hardness of 12Cr8Mo ferritic steels irradiated by Ni ions
ELSEVIER
Journal of Nuclear Materials 225 (1995) 187-191
Hardness of 12Cr-8Mo ferritic steels irradiated by Ni ions J. Ohta a, T. Ohmura a, K. Kako b, M. Tokiwai b, T. Suzuki a a Institute of Industrial Science, Universityof Tokyo, Roppongi, Minato-ku, Tokyo 106, Japan Central Research Institute of Electric PowerIndustry, Komae-shi, Tokyo 20l, Japan
Abstract
12Cr-8Mo and 12Cr-8Mo-0.1Y20 3 ferritic steels were irradiated with 4-MeV Ni 3÷ ions up to 300 dpa at 525°C. Microstructural evolution was examined by transmission electron microscopy (TEM) and mechanical properties were evaluated with a depth-sensing ultra-low load indentation hardness tester at room temperature with a maximum load of 1 gf. Effects of aging at 650°C for 115 h and heat treatment at 525°C for 50 h were also investigated. TEM observations reveal that these steels exhibit no void swelling in the present irradiation condition. Aging and heat treatment induces precipitation of Laves phase and ion-irradiation enhances precipitation. The induction and enhancement of precipitation strengthened the specimens.
1. Introduction
Because ferritic-martensitic steels are resistant to void swelling, they are used as wrapper tube materials in fast breeder reactors [1-8]. They also have good mechanical strength, creep resistance and a lower thermal expansion coefficient than austenitic stainless steels under 600°C [9-11]. However, their comparatively low creep strength above 600"C has restricted their application to cladding materials. Tokiwai et al. [12] have shown that newly developed 12Cr-8Mo ferritic steels, which are strengthened by (Fe,Cr) 2 Mo intermetallic compound precipitates (Laves phase), have twice the creep rupture strength of 12Cr1MoVW steel (HT-9) at 650°C. Addition of Y203 refines the grain size and further increases the creep rupture strength. They suggest the use of these steels as cladding material in metal-fueled fast breeders. Possible changes of microstructure and mechanical properties under irradiation remain a major concern. Heavy ion irradiation can produce high damage levels in very short periods of time. It has been utilized to study microstructure evolution under irradiation. However, evaluation of mechanical properties is limited because the damage is localized at the vicinity of the irradiated surface (typically < 1 ~tm) [13]. Zinkle and Oliver [14] have demonstrated that a depth-sensing ultra-low load indentation technique [15-17] can be
used to measure the mechanical properties of ion irradiated materials. In this work, we have observed microstructures and measured the ultra-microhardness of the 12Cr-8Mo steels irradiated by Ni ions.
2. Experimental 12Cr-8Mo and 12Cr-8Mo-0.1Y20 3 steels were manufactured by powder-metallurgy and mechanical alloying, respectively. 12Cr-8Mo steels were solution treated (ST) 1 h at 1050°C followed by 15% cold work. 12Cr-8Mo-0.1Y20 3 steels were solution treated 1 h at 1100~C and were not cold worked. Specimens which were solution treated and aged 115 h at 650°C (STA) were also prepared for both steels. Details are described elsewhere [12]. Discs of 0.2 mm thickness and 3mm diameter were cut. After electropolishing, the discs were irradiated with 4 MeV Ni 3+ ions to 300 dpa (in the peak damage region) at 525°C in the High Fluence Irradiation Facility, University of Tokyo, at a displacement damage rate of 1.7× 10 -3 d p a / s . According to calculations using the TRIM85 computer code [18], the damage is localized to a depth of 1.1 I~m from the irradiated surface. Transmission electron microscopy (TEM) foils were prepared by removing approximately 0.9 p.m from the irradiated surface and were backthinned by jet thinning to perforation. They
0022-3115/95/$09.50 © 1995 Elsevier Science B.V. All rights reserved SSDI 0022 -3115(95)00016-X
188
J. Ohta et al. /Journal of Nuclear Materials 225 (1995) 187-191
were examined in a JEM 200CX at 200 kV. At the depth of the TEM observations, thc damage level is 200 dpa and the concentration of Ni is 1 wt%. Effects of doping of Ni ions up to 1% on microstructures and mechanical properties are negligible [19]. The mechanical properties werc evaluated with a depth-sensing ultra-low load indentation hardness tester at room temperaturc with a maximum load of 1 gf at a loading rate of 0.1 gf/s. Inamura et al. [20] showed that for some metals the load L and indent depth d have the relationship L/d = Ad + B,
where A and B are dependent on materials but independent of load and indent depth. They also showed that A is proportional to the Vickers hardness and defined the ultra-microhardness H as H = CA, where C is an experimentally determined coefficient. We used this definition of hardness. A total of 10 indents were made in each of the irradiated and the unirradiated area in the same specimen. Because the indent depth was less than 300 nm, only the hardness of the damaged layer was obtained.
3. Results 3.1. Microstructure
Figs. 1-4 show electron micrographs of 12Cr-8Mo (ST and STA) and 12Cr-SMo-0.1Y20 3 (ST and STA) steels. The 12Cr-8Mo steel has a high density of dislocations, which are introduced by the cold work before aging (Fig. la). In a control specimen given the same thermal aging treatments as an irradiated specimen, viz. 525°C for 50 h, fine precipitation is induced (Fig. lb). Ion irradiation increases the population of precipitates (Fig. lc). After aging, precipitates exist in an as-received specimen and are coarsened by the thermal aging as well treatment as the irradiation (Fig. 2). In 12Cr-8Mo-0.1Y20 3 steels, grains are refined by dispersing Y203 . Precipitation along grain boundaries is observed. A high density of dislocations is observed in the unaged specimen (Fig. 3a). The effect of the thermal aging treatment and ion irradiation are similar to the effect in 12Cr-8Mo steels. That is, in an unaged specimen, fine precipitation was produced by the thermal aging treatment (Fig. 3b) and the population of precipitates was incrased by ion-irradiation (Fig. 3c). In an aged specimen, precipitates exist initially (Fig. 4a). As in 12Cr-8Mo steels, they were coarsened by both the thermal aging treatment and ion irradiation (Figs. 4b and c). As shown in Fig. 5, in an unaged specimen (Figs. 1 and 3) ion irradiation increases the population of pre-
Fig. 1. Electron micrographs of solution treated 12Cr-8Mo steels: (a) as-received, (b) control given a thermal aging treatment similar to that of the irradiated specimen and (c) irradiated.
cipitates. On the other hand, in an aged specimen (Figs. 2 and 4), ion irradiation increases the size of the precipitates. This suggests that ion irradiation primarily enhances nucleation or coarsening depending on whether precipitates are present before irradiation or not. Voids have never been observed in ion irradiated specimens. This means that these ferritic steels have high rcsistance to void swelling.
189
J. Ohta et al. /Journal of Nuclear Materials 225 (1995) 187-191 3.2. Ultra-microhardness
(8.) ,i.
As shown in Fig. 6, thermal aging hardens as-received 12Cr-8Mo steels by ~ 25%. The thermal aging treatment hardens an unaged 12Cr-8Mo steel (ST) by ~ 25% while hardening is substantially negligible in an aged specimen (STA). Ion irradiation further hardens thermally aged specimens by another ~ 40%. The hardness of as-received 12Cr-8Mo-0.1YzO3
~(b) /
llP-
,
?
.'_,,~111"
Fig. 3. Electron micrographs of solution treated 12Cr-8MoY203 steels: (a) as-received, (b) control given a thermal aging treatment similar to that of the irradiated specimen and (c) irradiated.
Fig. 2. Electron micrographs of solution treated and aged 12Cr-8Mo steels: (a) as-received, (b) control given a thermal aging treatment similar to that of the irradiated specimen and (c) irradiated.
steels is similar to that of 12Cr-8Mo steels. The thermal aging treatment hardens 12Cr-8Mo-0.1Y203 steels by ~ 40%. The hardening is larger than that observed in 12Cr-8Mo steels. Like 12Cr-8Mo steels, irradiation further hardens the thermally aged specimen. The hardening is partly due to the precipitation induced by the aging and enhancement of precipitation by the thermal aging treatment and the ion irradiation.
190
J. Ohta et al. /Journal of Nuclear Materials 225 (1995) 187-191
4. Discussion
[ as-recieved •
4.1. Swelling resistance
Void formation occurs in ferritic and martensitic alloys over a relatively narrow temperature range from around 400 to 450°C for neutron irradiations [3,4,21,22] and from about 500 to 600°C for charged particle irradiations [23]. Thus, it is expected that voids can be formed at the temperature at which the present ion
]control
25
•
E
o~ Q -
irradiated
20
X v
~,
15
IL
a
z
5
12Cr-8Mo 12Cr-SMo (ST)
(STA)
12Cr-8Mo- 12Cr-8Mo0.1Y203 0.1Y203 (ST) (STA)
Fig. 5. Precipitates in 12Cr-8Mo (solution treated (ST) and solution treated and aged (STA)) and 12Cr-8Mo-0.1YzO 3 steels (solution treated (ST) and solution treated and aged (STA)). Control is given a thermal aging treatment similar to that of the irradiated specimen.
irradiation was performed, 525°C. However, voids are not observed. This implies that 12Cr-8Mo and 12Cr8Mo-0.1Y20 3 steels have a high resistance to void swelling. Mechanisms of high resistance to void swelling in ferritic steels are described in recent reviews; intrinsic lower dislocation bias, partitioning of excess interstitials and vacancies to dislocations, defect solute trapping leading to increased recombination, fine scale subgrain structure, and so on [24-26]. In addition to these mechanisms, the extremely low void swelling of these steels suggest that another mechanism should
[ as-recieved • 800 ~. (3.
.~
[ ] control •
irradiated
6O0
400
b 200
0 12Cr-8Mo 12Cr-8Mo 12Cr-8Mo- 12Cr-8Mo0.1Y203 0.1Y203 (ST) (STA) (ST) (STA) Fig. 4. Electron micrographs of ,solution treated and aged 12Cr-8Mo-Y20 3 steels: (a) as-received, (b) control given a thermal aging treatment similar to that of the irradiated specimen and (c) irradiated.
Fig. 6. Ultra-microhardness of 12Cr-8Mo (solution treated (ST) and solution-treated-and-aged (STA)) and 12Cr-8Mo0.1Y20 3 steels (solution-treated (ST) and solution-treatedand-aged (STA)). Control is given a thermal aging treatment similar to that of the irradiated specimen.
J. Ohta et al. /Journal of Nuclear Materials 225 (1995) 187-191 also operate: the high density of precipitates should provide an additional significant sink for point defects. 4.2. Ultra-microhardness Peterson [27] found that changes in the impact behavior of 9 C r - I M o V N b and 1 2 C r - I M o V W steels aged for 1000 h correlated with the appearance of a high density of fine Laves phase particles at 500°C, and with the total amount of precipitate (Laves + carbides) at 600°C. Hardening of 1 2 C r - 8 M o and 1 2 C r - 8 M o 0.1Y20 3 steels should also be related to the precipitate reactions that occur during aging and ion irradiation. Kako and Tokiwai [28] have investigated the effect of aging on the precipitation and Vickers hardness in the steels. In the same aging condition, the hardness values of both the 1 2 C r - 8 M o and the 1 2 C r - 8 M o - 0 . 1 Y 2 0 3 steels are similar although the precipitates in the 1 2 C r - 8 M o steels are denser and finer than those in the 1 2 C r - 8 M o - 0 . 1 Y 2 0 3 steels. It implies that the hardness of the steels cannot correlate with only the number density of precipitates. The manufacturing process and chemical composition must also affect the hardness. For the irradiated specimens, we should consider additional hardening by radiation-induced defect clusters, which are not detected directly by T E M . In F c - C r alloys, hardening caused by extremely small defect clusters is reported by Suganuma and Kayano [29]. Because of these various factors which can affect the hardness, we cannot find a quantitative correlation between the hardness and the microstructure.
5. Conclusion 1 2 C r - 8 M o and 1 2 C r - 8 M o - 0 . 1 Y 2 0 3 steels do not exhibit void swelling during Ni ion irradiation up to 300 dpa at 525°C. Ni ion irradiation hardens these steels by e n h a n c e m e n t of precipitation.
References [1] K.Q. Bagley, E.A. Little, V. l.~vy, A. Alamo, K. Ehrlich, K. Anderko and A. Calzahini, Materials for Nuclear
191
Reactor Core Applications, vol. 2 (BNES, London, 1987) pp. 37-46. [2] E.A. Little and D.A. Stow, J. Nucl. Mater. 87 (1979) 25. [3] D.S. Geiles, J. Nuci. Mater. 122&123 (1984) 207. [4] D.S. Gelles, J. Nucl. Mater. 148 (1987) 136. [5] D. Gilbon, J.L. S6ran R. Cauvin, A. Fissolo, A. Alamo, F. LeNaour and V. L~vy, ASTM-STP 1046 (1989) 5. [6] J.L. S6ran, V. L6vy, P. Dubuisson, D. Gilbon, A. Maillard, A. Fissolo, H. Touron, R. Cauvin, A. Chalony and E. Le Boulbin, ASTM-STP 1125 (19921 1209. [71 P. Dubisson, D. Gilbon and J.L. S6ran, J. Nucl. Mater. 205 (1993) 178. [8] E.A. Little, J. Nucl. Mater. 206 (1993) 324. [9] M.L. Grossbeck, J.M. Vitek and K.C. Liu, J. Nucl. Mater. 141-143 (1986) 966. [10] L.A. James, J. Nucl. Mater. 149 (1987) 138. [11] J.W. Davis, J. Nucl. Mater. 122&123 (1984) 51. [12] M. Tokiwai, M. Horie, K. Kako and M. Fujiwara, J. Nucl. Mater. 204 (1993) 56. [13] G.E. Lucas, Metall. Trans. 21A (1990) 1105. [14] S.J. Zinklc and W.C. Oliver, J. Nucl. Mater. 141-143 (1986) 548. [15] J.B. Pethica, R. Hutchings and W.C. Oliver, Philos. Mag. A48 (1983) 593. [16] W.C. Oliver, R. Hutchings and J.B. Pethica, ASTM-STP 889 (1986) 90. [17] M.F. Doerner and W.D. Nix, J. Mater. Res. 1 (1986) 601. [18] K.F. Ziegler, J.P. Biersack and U. Littmark, The Stopping and Range of Ions in Solids (Pergamon, New York. 1985) pp. 109-140. [19] K. Kako, unpublished. [20] M. lnamura and T. Suzuki, Seisan Kenkyu 42 (1990) 257 (in Japanese). [21] P. Dubisson, D. Gilbon and J.L. S6ran, J. Nucl. Mater. 205 (1993) 178. [22] D.S. Gelles, J. Nucl. Mater. 103&104 (1981) 975. [23] K. Farrell and E. Lee, ASTM-STP 870 (1985) 383. [24] P.J. Maziaz and R.L. Klueh, ASTM-STP 1046 (1989) 35. [25] E.A. Little, Materials for Nuclear Reactor Core Applications, vol. 2 (BNES, London, 1987) pp. 47-55. [26] G.R. Odette, J..NucL Mater. 155-157 (1988) 921. [27] D.T. Petersen, US Department of Energy Report DOE/ER-0045/5 (Office of Fusion Energy, Washington, DC, 1980) p. 212. [28] K. Kako and M. Tokiwai, CRIEPI Report T 92094 (1994) (in Japanese). [29] K. Suganuma and H. Kayano, J. Nucl. Mater. 118 (1983) 234.
Journal of Nuclear Materials 225 (1995) 187-191
Hardness of 12Cr-8Mo ferritic steels irradiated by Ni ions J. Ohta a, T. Ohmura a, K. Kako b, M. Tokiwai b, T. Suzuki a a Institute of Industrial Science, Universityof Tokyo, Roppongi, Minato-ku, Tokyo 106, Japan Central Research Institute of Electric PowerIndustry, Komae-shi, Tokyo 20l, Japan
Abstract
12Cr-8Mo and 12Cr-8Mo-0.1Y20 3 ferritic steels were irradiated with 4-MeV Ni 3÷ ions up to 300 dpa at 525°C. Microstructural evolution was examined by transmission electron microscopy (TEM) and mechanical properties were evaluated with a depth-sensing ultra-low load indentation hardness tester at room temperature with a maximum load of 1 gf. Effects of aging at 650°C for 115 h and heat treatment at 525°C for 50 h were also investigated. TEM observations reveal that these steels exhibit no void swelling in the present irradiation condition. Aging and heat treatment induces precipitation of Laves phase and ion-irradiation enhances precipitation. The induction and enhancement of precipitation strengthened the specimens.
1. Introduction
Because ferritic-martensitic steels are resistant to void swelling, they are used as wrapper tube materials in fast breeder reactors [1-8]. They also have good mechanical strength, creep resistance and a lower thermal expansion coefficient than austenitic stainless steels under 600°C [9-11]. However, their comparatively low creep strength above 600"C has restricted their application to cladding materials. Tokiwai et al. [12] have shown that newly developed 12Cr-8Mo ferritic steels, which are strengthened by (Fe,Cr) 2 Mo intermetallic compound precipitates (Laves phase), have twice the creep rupture strength of 12Cr1MoVW steel (HT-9) at 650°C. Addition of Y203 refines the grain size and further increases the creep rupture strength. They suggest the use of these steels as cladding material in metal-fueled fast breeders. Possible changes of microstructure and mechanical properties under irradiation remain a major concern. Heavy ion irradiation can produce high damage levels in very short periods of time. It has been utilized to study microstructure evolution under irradiation. However, evaluation of mechanical properties is limited because the damage is localized at the vicinity of the irradiated surface (typically < 1 ~tm) [13]. Zinkle and Oliver [14] have demonstrated that a depth-sensing ultra-low load indentation technique [15-17] can be
used to measure the mechanical properties of ion irradiated materials. In this work, we have observed microstructures and measured the ultra-microhardness of the 12Cr-8Mo steels irradiated by Ni ions.
2. Experimental 12Cr-8Mo and 12Cr-8Mo-0.1Y20 3 steels were manufactured by powder-metallurgy and mechanical alloying, respectively. 12Cr-8Mo steels were solution treated (ST) 1 h at 1050°C followed by 15% cold work. 12Cr-8Mo-0.1Y20 3 steels were solution treated 1 h at 1100~C and were not cold worked. Specimens which were solution treated and aged 115 h at 650°C (STA) were also prepared for both steels. Details are described elsewhere [12]. Discs of 0.2 mm thickness and 3mm diameter were cut. After electropolishing, the discs were irradiated with 4 MeV Ni 3+ ions to 300 dpa (in the peak damage region) at 525°C in the High Fluence Irradiation Facility, University of Tokyo, at a displacement damage rate of 1.7× 10 -3 d p a / s . According to calculations using the TRIM85 computer code [18], the damage is localized to a depth of 1.1 I~m from the irradiated surface. Transmission electron microscopy (TEM) foils were prepared by removing approximately 0.9 p.m from the irradiated surface and were backthinned by jet thinning to perforation. They
0022-3115/95/$09.50 © 1995 Elsevier Science B.V. All rights reserved SSDI 0022 -3115(95)00016-X
188
J. Ohta et al. /Journal of Nuclear Materials 225 (1995) 187-191
were examined in a JEM 200CX at 200 kV. At the depth of the TEM observations, thc damage level is 200 dpa and the concentration of Ni is 1 wt%. Effects of doping of Ni ions up to 1% on microstructures and mechanical properties are negligible [19]. The mechanical properties werc evaluated with a depth-sensing ultra-low load indentation hardness tester at room temperaturc with a maximum load of 1 gf at a loading rate of 0.1 gf/s. Inamura et al. [20] showed that for some metals the load L and indent depth d have the relationship L/d = Ad + B,
where A and B are dependent on materials but independent of load and indent depth. They also showed that A is proportional to the Vickers hardness and defined the ultra-microhardness H as H = CA, where C is an experimentally determined coefficient. We used this definition of hardness. A total of 10 indents were made in each of the irradiated and the unirradiated area in the same specimen. Because the indent depth was less than 300 nm, only the hardness of the damaged layer was obtained.
3. Results 3.1. Microstructure
Figs. 1-4 show electron micrographs of 12Cr-8Mo (ST and STA) and 12Cr-SMo-0.1Y20 3 (ST and STA) steels. The 12Cr-8Mo steel has a high density of dislocations, which are introduced by the cold work before aging (Fig. la). In a control specimen given the same thermal aging treatments as an irradiated specimen, viz. 525°C for 50 h, fine precipitation is induced (Fig. lb). Ion irradiation increases the population of precipitates (Fig. lc). After aging, precipitates exist in an as-received specimen and are coarsened by the thermal aging as well treatment as the irradiation (Fig. 2). In 12Cr-8Mo-0.1Y20 3 steels, grains are refined by dispersing Y203 . Precipitation along grain boundaries is observed. A high density of dislocations is observed in the unaged specimen (Fig. 3a). The effect of the thermal aging treatment and ion irradiation are similar to the effect in 12Cr-8Mo steels. That is, in an unaged specimen, fine precipitation was produced by the thermal aging treatment (Fig. 3b) and the population of precipitates was incrased by ion-irradiation (Fig. 3c). In an aged specimen, precipitates exist initially (Fig. 4a). As in 12Cr-8Mo steels, they were coarsened by both the thermal aging treatment and ion irradiation (Figs. 4b and c). As shown in Fig. 5, in an unaged specimen (Figs. 1 and 3) ion irradiation increases the population of pre-
Fig. 1. Electron micrographs of solution treated 12Cr-8Mo steels: (a) as-received, (b) control given a thermal aging treatment similar to that of the irradiated specimen and (c) irradiated.
cipitates. On the other hand, in an aged specimen (Figs. 2 and 4), ion irradiation increases the size of the precipitates. This suggests that ion irradiation primarily enhances nucleation or coarsening depending on whether precipitates are present before irradiation or not. Voids have never been observed in ion irradiated specimens. This means that these ferritic steels have high rcsistance to void swelling.
189
J. Ohta et al. /Journal of Nuclear Materials 225 (1995) 187-191 3.2. Ultra-microhardness
(8.) ,i.
As shown in Fig. 6, thermal aging hardens as-received 12Cr-8Mo steels by ~ 25%. The thermal aging treatment hardens an unaged 12Cr-8Mo steel (ST) by ~ 25% while hardening is substantially negligible in an aged specimen (STA). Ion irradiation further hardens thermally aged specimens by another ~ 40%. The hardness of as-received 12Cr-8Mo-0.1YzO3
~(b) /
llP-
,
?
.'_,,~111"
Fig. 3. Electron micrographs of solution treated 12Cr-8MoY203 steels: (a) as-received, (b) control given a thermal aging treatment similar to that of the irradiated specimen and (c) irradiated.
Fig. 2. Electron micrographs of solution treated and aged 12Cr-8Mo steels: (a) as-received, (b) control given a thermal aging treatment similar to that of the irradiated specimen and (c) irradiated.
steels is similar to that of 12Cr-8Mo steels. The thermal aging treatment hardens 12Cr-8Mo-0.1Y203 steels by ~ 40%. The hardening is larger than that observed in 12Cr-8Mo steels. Like 12Cr-8Mo steels, irradiation further hardens the thermally aged specimen. The hardening is partly due to the precipitation induced by the aging and enhancement of precipitation by the thermal aging treatment and the ion irradiation.
190
J. Ohta et al. /Journal of Nuclear Materials 225 (1995) 187-191
4. Discussion
[ as-recieved •
4.1. Swelling resistance
Void formation occurs in ferritic and martensitic alloys over a relatively narrow temperature range from around 400 to 450°C for neutron irradiations [3,4,21,22] and from about 500 to 600°C for charged particle irradiations [23]. Thus, it is expected that voids can be formed at the temperature at which the present ion
]control
25
•
E
o~ Q -
irradiated
20
X v
~,
15
IL
a
z
5
12Cr-8Mo 12Cr-SMo (ST)
(STA)
12Cr-8Mo- 12Cr-8Mo0.1Y203 0.1Y203 (ST) (STA)
Fig. 5. Precipitates in 12Cr-8Mo (solution treated (ST) and solution treated and aged (STA)) and 12Cr-8Mo-0.1YzO 3 steels (solution treated (ST) and solution treated and aged (STA)). Control is given a thermal aging treatment similar to that of the irradiated specimen.
irradiation was performed, 525°C. However, voids are not observed. This implies that 12Cr-8Mo and 12Cr8Mo-0.1Y20 3 steels have a high resistance to void swelling. Mechanisms of high resistance to void swelling in ferritic steels are described in recent reviews; intrinsic lower dislocation bias, partitioning of excess interstitials and vacancies to dislocations, defect solute trapping leading to increased recombination, fine scale subgrain structure, and so on [24-26]. In addition to these mechanisms, the extremely low void swelling of these steels suggest that another mechanism should
[ as-recieved • 800 ~. (3.
.~
[ ] control •
irradiated
6O0
400
b 200
0 12Cr-8Mo 12Cr-8Mo 12Cr-8Mo- 12Cr-8Mo0.1Y203 0.1Y203 (ST) (STA) (ST) (STA) Fig. 4. Electron micrographs of ,solution treated and aged 12Cr-8Mo-Y20 3 steels: (a) as-received, (b) control given a thermal aging treatment similar to that of the irradiated specimen and (c) irradiated.
Fig. 6. Ultra-microhardness of 12Cr-8Mo (solution treated (ST) and solution-treated-and-aged (STA)) and 12Cr-8Mo0.1Y20 3 steels (solution-treated (ST) and solution-treatedand-aged (STA)). Control is given a thermal aging treatment similar to that of the irradiated specimen.
J. Ohta et al. /Journal of Nuclear Materials 225 (1995) 187-191 also operate: the high density of precipitates should provide an additional significant sink for point defects. 4.2. Ultra-microhardness Peterson [27] found that changes in the impact behavior of 9 C r - I M o V N b and 1 2 C r - I M o V W steels aged for 1000 h correlated with the appearance of a high density of fine Laves phase particles at 500°C, and with the total amount of precipitate (Laves + carbides) at 600°C. Hardening of 1 2 C r - 8 M o and 1 2 C r - 8 M o 0.1Y20 3 steels should also be related to the precipitate reactions that occur during aging and ion irradiation. Kako and Tokiwai [28] have investigated the effect of aging on the precipitation and Vickers hardness in the steels. In the same aging condition, the hardness values of both the 1 2 C r - 8 M o and the 1 2 C r - 8 M o - 0 . 1 Y 2 0 3 steels are similar although the precipitates in the 1 2 C r - 8 M o steels are denser and finer than those in the 1 2 C r - 8 M o - 0 . 1 Y 2 0 3 steels. It implies that the hardness of the steels cannot correlate with only the number density of precipitates. The manufacturing process and chemical composition must also affect the hardness. For the irradiated specimens, we should consider additional hardening by radiation-induced defect clusters, which are not detected directly by T E M . In F c - C r alloys, hardening caused by extremely small defect clusters is reported by Suganuma and Kayano [29]. Because of these various factors which can affect the hardness, we cannot find a quantitative correlation between the hardness and the microstructure.
5. Conclusion 1 2 C r - 8 M o and 1 2 C r - 8 M o - 0 . 1 Y 2 0 3 steels do not exhibit void swelling during Ni ion irradiation up to 300 dpa at 525°C. Ni ion irradiation hardens these steels by e n h a n c e m e n t of precipitation.
References [1] K.Q. Bagley, E.A. Little, V. l.~vy, A. Alamo, K. Ehrlich, K. Anderko and A. Calzahini, Materials for Nuclear
191
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