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Microstructure Characterization of Ion-irradiated Ferritic/Martensitic HT9 Steel - Volume 23 Issue S1 - Djamel Kaoumi, Ce Zheng
Microstructure Characterization of Ion-irradiated Ferritic/Martensitic HT9 Steel
Djamel Kaoumi1, Ce Zheng1
1 Department of Nuclear Engineering, North Carolina State University, Raleigh, 27607, NC, USA
HT9 is a 12Cr Ferritic/Martensitic (F/M) steel considered as a promising candidate for structural
and cladding applications in Generation IV reactors [1]. The chemical composition of the alloy is given
in Table 1. The harsh service conditions in Gen IV reactors require that the microstructural response to
irradiation of the candidate structural alloys be investigated and understood to qualify them. For that
matter, a series of ion irradiations were done. Bulk HT9 specimens were irradiated using 5 MeV Fe++
ions to 20 displacements per atom (dpa) at 600 nm depth with a dose rate approximately to 5×10-4 dpa/s,
at irradiation temperatures of 420, 440 and 470°C (with a variation of ± 5°C). The temperature was
monitored using an infrared camera and four attached Type J thermocouples. For post-irradiation
characterization, TEM specimens were firstly prepared by the FIB lift-out method using a FEI Quanta
focused ion beam (FIB) instrument. ChemiSTEM characterizations were then conducted on FIB laminas
using a FEI Titan 80-300 probe aberration corrected microscope. ChemiSTEM characterization was also
conducted on as-received HT9 prior to ion irradiation. Only pre-existed M23C6 type carbides and V-rich
nitride precipitates were observed in the as-received condition. In contrast, Ni-Si-Mn rich precipitates
(also known as G phase) were found in HT9 irradiated to 20 dpa at 420, 440 and 470°C, as shown in
Figure 1. Radiation-induced Ni segregation was also observed at grain boundaries, which is highlighted
by white arrows in Figure 2. In addition, the G phase precipitates were found to nucleate
heterogeneously along lath grain boundaries, as indexed by red arrows in Figure 2. The observed results
indicated that, under self-ion irradiation, alloying elements such as Ni, Si and Mn segregate at defect
sinks, which become thus favourable nucleation sites and promote the radiation-induced G phase
precipitation. While radiation induced precipitation and segregation in neutron irradiated F/M HT9 have
been widely reported in the literature, similar investigations under ion irradiation have been more scarce
[2,3,4]. In fact, this study serves to generate baseline data on ion irradiation effects on F/M HT9 in an
effort to learn how to more accurately choose ion-irradiation experimental conditions to emulate the
irradiated microstructures and effects observed under neutron irradiation.
Ion irradiations were also carried out in the same alloy at similar temperatures in-situ in a TEM
using 1MeV Kr++ ions so that the microstructure characterized in-situ in the TEM can be compared
with the microstructure achieved on the same alloys using self-ion irradiation on bulk samples. The
focus of the comparison is put on the size and density of dislocation loops induced by irradiation, as well
as dislocation loop burgers vector determination. The in-situ experiments provide data on the kinetics of
irradiation induced defect formation and evolution, and on the damage spatial correlation with the pre-
existing microstructure, and thus can help understand how the microstructures observed ex-situ in the
bulk samples have developed for, in these latter cases, only snapshots are available at the limited doses.
By comparing the ex-situ and in-situ irradiation it is also possible to substantiate the free surface effect
on the radiation induced microstructure. The presentation will also report such comparison.
References:
[1] R. L. Klueh, A. T. Nelson, Journal of Nuclear Materials, 371 (2007) p. 37
[2] J. J. Kai, G. L. Kulcinshi, Journal of Nuclear Materials, 175 (1990) p. 227
[3] E. Getto et al, Journal of Nuclear Materials, 480 (2016) p. 159
[4] E. Getto et al, Journal of Nuclear Materials, 484 (2017) p.193
2214
doi:10.1017/S1431927617011734 Microsc. Microanal. 23 (Suppl 1), 2017
© Microscopy Society of America 2017
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Fe
Cr
Mo
Ni
Mn
W
V
Si
C
N
P
S
84.89
11.8
1.03
0.51
0.5
0.5
0.33
0.21
0.21
0.01
0.008
0.003
Table 1. As-received HT9 chemical composition (wt.%)
Figure 1. ChemiSTEM maps showed Ni, Si and Mn enrichment (associated with Fe depletion) in HT9
self-ion irradiated to 20dpa at 420, 440 and 470°C. It indicates the formation of Ni, Si, Mn-rich
precipitates in HT9 under irradiation.
Figure 2. STEM mode bright field micrograph (a) and corresponding Ni map (b) of HT9 self-ion
irradiated to 20 dpa at 440°C; Ni segregated to grain boundaries (indexed by white arrows) and some of
Ni, Si, Mn-rich precipitates nucleated heterogeneously along lath grain boundaries (indexed by red
arrows).
2215Microsc. Microanal. 23 (Suppl 1), 2017
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Article
The ferritic/martensitic steel HT9 was irradiated in the BOR-60 reactor at 650, 690 and 730 K (377, 417 and 457ºC) to doses between ~14.6-18.6 displacements per atom (dpa). Irradiated samples were comprehensively characterized using analytical scanning/transmission electron microscopy and atom probe tomography, with emphasis on the influence of irradiation temperature on microstructure evolution. Mn/Ni/Si-rich (G-phase) and Cr-rich (αʹ) precipitates were observed within martensitic laths and at various defect sinks at 650 and 690 K (377 and 417ºC). For both G-phase and αʹ precipitates, the number density decreased while the size increased with increasing temperature. At 730 K (457ºC), within martensitic laths, a very low density of large G-phase precipitates nucleating presumably on dislocation lines was observed. No αʹ precipitates were observed at this temperature. Both a <100> and a/2 <111> type dislocation loops were observed, with the a <100> type being the predominant type at 650 and 690 K (377 and 417ºC). On the contrary, very few dislocation loops were observed at 730 K (457ºC), and the microstructure was dominated by a/2 <111> type dislocation lines (i.e., dislocation network) at this temperature. Small cavities (diameter < 2 nm) were observed at all three temperatures, whereas large cavities (diameter > 2 nm) were observed only at 690 K (417ºC), resulting in a bimodal cavity size distribution at 690 K (417ºC) and a unimodal size distribution at 650 and 730 K (377 and 457ºC). The highest swelling (%) was observed at 690 K (417ºC), indicating that the peak of swelling happens between 650 and 730 K (377 and 457ºC).
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Understanding the void swelling and phase evolution of reactor structural materials at very high damage levels is essential to maintaining safety and longevity of components in Gen IV fast reactors. A combination of ion irradiation and modeling was utilized to understand the microstructure evolution of ferritic-martensitic alloy HT9 at high dpa. Self-ion irradiation experiments were performed on alloy HT9 to determine the co-evolution of voids, dislocations and precipitates up to 650 dpa at 460 °C. Modeling of microstructure evolution was conducted using the modified Radiation Induced Microstructure Evolution (RIME) model, which utilizes a mean field rate theory approach with grouped cluster dynamics. Irradiations were performed with 5 MeV raster-scanned Fe²⁺ ions on samples pre-implanted with 10 atom parts per million He. The swelling, dislocation and precipitate evolution at very high dpa was determined using Analytical Electron Microscopy in Scanning Transmission Electron Microscopy (STEM) mode. Experimental results were then interpreted using the RIME model. A microstructure consisting only of dislocations and voids is insufficient to account for the swelling evolution observed experimentally at high damage levels in a complicated microstructure such as irradiated alloy HT9. G phase was found to have a minimal effect on either void or dislocation evolution. M2X plays two roles; a variable biased sink for defects, and as a vehicle for removal of carbon from solution, thus promoting void growth. When accounting for all microstructure interactions, swelling at high damage levels is a dynamic process that continues to respond to other changes in the microstructure as long as they occur.
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The void swelling and microstructure evolution of ferritic-martensitic alloys HT9, T91 and T92 were characterized following irradiation with Fe þþ ions at 460 C to damage levels of 75e650 displacements per atom with 10 atom parts per million pre-implanted helium. Steady state swelling rate of 0.033%/ dpa was determined for HT9, the least swelling resistant alloy, and 0.007%/ dpa in T91. In T91, resistance was due to suppression of void nucleation. Swelling resistance was greatest in T92, with a low density (~1 Â 10 20 m À3) of small voids that had not grown appreciably, indicating suppression of nucleation and growth. Additional heats of T91 indicated that alloy composition was not the determining factor of swelling resistance. Carbon and chromium-rich M 2 X precipitates formed at 250 dpa and were correlated with decreased nucleation in T91 and T92, but did not affect void growth in HT9. Dislocation and G-phase microstructure evolution was analyzed up to 650 dpa in HT9. Published by Elsevier B.V.
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Two sets of HT-9 specimens were irradiated with 14 MeV nickel ions. In which, one set of specimens were irradiated in the temperature range from 300 to 600°C and up to a peak dose level of 200 dpa and the other set of specimens were uniformly preimplanted with 100 appm of helium and then were irradiated at temperatures of 400, 500, and 600°C up to 60 dpa. The results indicated that there was virtually no void swelling at all in no helium specimens and even with 100 appm helium pre-implantation the maximum local swelling was found to be 0.1% in the specimens irradiated at 500°C to 60 dpa. This result furtherly demonstrated the superior void swelling resistance of HT-9 ferritic steel and it is also concluded that free gas atoms are essential to the formation of cavities in heavy-ion irradiated HT-9. Other microstructural features resulting from the irradiation were also examined in detail, including dislocation loops and precipitates.
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E. Getto et al, Journal of Nuclear Materials, 484 (2017) p.193