J. Plasma Fusion Res. SERIES, Vol. 11 (2015)

Effects of Ta and C on irradiation hardening of austenitic stainless steel
Yasuyuki SAKAI1, Yutaka KURIBAYASHI1, Kiyohiro YABUUCHI1, Akira HASEGAWA1
Shuhei NOGAMI1, Junya KANEDA2, Naoto SHIGENAKA2,Department of Quantum Science and Energy Engineering, Tohoku University, 6-6-01-2 Aramaki-aza-Aoba, Aobaku,Sendai, Miyagi, 980-8579, Japan2
Hitachi-GE Nuclear Energy, 3-1-1 Saiwai-cho, Hitachi, Ibaraki, 317-0073, Japan1(Received: 17 September 2014 / Accepted: 19 January 2015)

Type 304L austenitic stainless steels are used for reactor internal components of light-water reactors and theirradiation assisted stress corrosion cracking (IASCC) is one of the degradation issues. We have developed IASCC
resistant stainless steels focusing on minor addition of Ta as an oversized element. In this study, the effect of Ta andC on the irradiation hardening of the alloys developed was investigated. 304L austenitic stainless steels with various
Ta and C contents were fabricated and characterized. These alloys were irradiated by 1 MeV H+ ion beam up todisplacement damage of 0.1, 0.3 and 1 dpa at 473K, 573K, 623K and 673K. In order to investigate the irradiation
hardening, Vickers hardness test was carried out and the microstructure was observed by transmission electron
microscopy. The irradiation hardening of the alloys was almost the same as that of 304L. It is expected that Ta andC addition alloy increase IASCC resistance due to reducing irradiation damage.
Keywords: irradiation assisted stress corrosion cracking, austenitic stainless steel, Ta addition, proton irradiation,
irradiation hardening, microstructure.

1. Introduction

intergranular corrosion resistance decreases. For thepurpose of preventing RIS, the oversized element
addition such as Hf, Zr, Ta and Nb to austenitic stainlesssteels has been suggested because the additionaloversized atoms dominantly trap vacancies and prevent
Cr diffusion to grain boundary [4].
We have tried to develop IASCC resistant stainlesssteels focusing on Ta as an oversized element. In thisstudy, the irradiation hardening of the steels underdevelopment are investigated by ion irradiation.

Austenitic stainless steels are used for reactor
internal components of light-water reactors (LWRs)
because of its excellent corrosion resistance, workability
and weldability. The reactor internal components play
important roles in supporting the fuel assemblies andcontrol rods, providing a flow path of cooling water andso on. These components are required to keep theirsoundness under the severe condition such as neutron
irradiation in high temperature and high pressure water.
Neutron irradiation causes various material propertychanges, such as irradiation hardening, embrittlement,swelling and irradiation assisted stress corrosion cracking(IASCC) [1, 2]. Especially, IASCC is one of thedegradation issues of LWR. It is believed thatone of theIASCC causes is radiation induced segregation (RIS) [3].
In the case of IASCC, depletion of Cr in the vicinity ofgrain boundaries due to the vacancy diffusion is the key
phenomenon. Vacancies introduced by neutron irradiation
diffuse to grain boundaries, exchanging the position
preferentially with oversized elements such as Cr or Mn.
As a result, Cr is depleted at grain boundary, and the2. Experimental

Type 304L stainless steels with addition of Ta and C
were used in this study: these were Low-Ta Low-C alloy,Middle-Ta Low-C alloy, Middle-Ta Middle-C alloy andHigh-Ta High-C alloy. The chemical compositions areshown in Table 1. The Ta/C ratio is respectively 0.86,2.47, 0.90 and 0.88. Disks of 3mm in diameter werepunched, and the disks were solution heat treated at1323K for 1 h in vacuum and were quenched in water.
Then, stabilizing heat treatment was carried out at 1173K
for 1 h in vacuum, and the specimens were quenched inTable 1 Chemical composition of specimens.: No unit for Ta/C.

yasuyuki.sakai@jupiter.qse.tohoku.ac.jp

83©2015 by The Japan Society of Plasma
Science and Nuclear Fusion Research

Y. Sakai et al., Effects of Ta and C on Irradiation Hardening of Austenitic Stainless Steelhardening to 0.3 dpa. Irradiation hardening was decreased
with increasing irradiation temperature.
The indentation depth of all tests ranged from 1.0 ~1.4 μm. If the deformation region was five times theindentation depth, the deformation region can beestimated to range from 5.0 ~ 7.2 μm. It is included in theirradiated region as shown in Fig. 1. Moreover, Yabuuchi
et al. discussed on the relationship between Vickeres
hardness and nanoindentation hardness with proton
irradiated stainless steels [7], and the test condition of thisstudy is the same. Therefore, the results of Vickeres
hardness test in this study were reasonable.

water. The average grain size after the heat treatment wasabout 27 μm. The specimen surface was mechanicalpolished and electropolished.
To introduce irradiation defects to the specimens,
ion irradiation was carried out using Dynamitron
accelerator at Tohoku University [5]. 1 MeV H+ was usedin this experiment. The displacement damage was 0.1, 0.3
and 1 dpa at the depth of 4 μm from the surface with theirradiation rate of 1.8 ൈ 10-5 dpa/s. Fig.1 shows thedisplacement damage distribution calculated by
SRIM2013 code [6]. Irradiation Temperature was 473K,573K, 623K and 673K, which was measured bythermography. These irradiation experiments wereconducted in a vacuum of less than 5.0 ൈ 10-6 Torr.
Vickers hardness test was carried out to measureirradiation hardening, where the irradiation hardening is adifference between the hardness after and before theirradiation. Test load was 9.8 ൈ 10-2 N, holding time was
15 sec, and test temperature was R.T.
In order to investigate effects of proton irradiation
on microstructure evolution, transmission electron
microscopy (TEM) and energy dispersive X-ray
spectroscope (EDS) observation were conducted using
JEM-2100F (JEOL) operating at 200kV. The thicknessesof the thin films were measured using convergent beamelectron diffraction method.

3.2 TEM observation
Fig. 4 shows the images of TEM observation before
irradiation. Precipitates were observed in High-Ta High-Cand Middle-Ta Middle-C alloys. Fig. 5 shows result ofEDS, (a) line analysis and (b)-(f) area analysis. As Ta was

3. Result and discussion

3.1 Vickers hardness test
Fig. 2 shows the displacement damage dependenceof irradiation hardening of specimens irradiated at 573K.
Irradiation hardening increased with increasing
displacement damage for all specimens. Irradiation
hardening behavior was almost the same between the Ta
added alloys and 304L stainless steel. Fig. 3 shows theirradiation temperature dependence of irradiation

Fig. 2 Irradiation hardening as a function ofdisplacement damage.

Fig. 1 Depth distribution of displacement
damage.

Fig. 3 Irradiation hardening as a function ofirradiation temperature.

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Y. Sakai et al., Effects of Ta and C on Irradiation Hardening of Austenitic Stainless Steel

Fig. 4 The TEM images of unirradiated samples.

Fig. 5 The images of EDS (a) line analysis and (b)-(f) area analysis of unirradiated High-Ta High-C
stainless steel and Ta addition alloys irradiated at 473K to673K. Dislocation loop size and number density as afunction of irradiation temperature were shown in Fig. 10and Fig. 11. The dislocation loop size increased withtemperature.

mainly detected from precipitates as shown in Fig. 5, theprecipitates were identified to be TaC.
TEM images of 304L stainless steel and Middle-Ta
Low-C alloy unirradiated and irradiated to 0.3 dpa and1dpa are shown in Fig. 6. Dislocation loops wereobserved in the specimens irradiated at 0.3 and 1 dpa.
Fig.7 and Fig. 8 show dislocation loop size and number
density dependence as a function of displacement damage.
Evolution of dislocation loop in Middle-Ta Low-C andLow-Ta Low-C alloy was suppressed relative to 304L
stainless steel. Fig. 9 shows TEM images of 304L

3.3 Application of Orowan model
In order to clarify the contribution of dislocation
loops to irradiation hardening, the increase in yield stress
was estimated by Orowan model [8] using the following
equation:85Y. Sakai et al., Effects of Ta and C on Irradiation Hardening of Austenitic Stainless Steel

Fig. 6 TEM micrograph of 304L and Middle-Ta Low-C alloy irradiated up to 1 dpa.

Fig. 8 Dislocation loop density vs. displacement
damage.

Fig. 7 Average dislocation loop size vs.
displacement damage.
οߪ௖௔௟௖ ൌ ‫ܾߤߙܯ‬ሺܰ݀ሻ଴Ǥହοߪ௘௫௣ ൌ ͵ǤͲ͵ο‫ܪ‬௏

(1)

(2)

The correlation between calculated yield stress οߪ௖௔௟௖ ,and experimental value οߪ௘௫௣ is shown in Fig. 12. Asshown in Fig. 12, the calculated yield stress, οߪ௖௔௟௖ , isnearly equal to the experimental value, οߪ௘௫௣ ,suggesting that the irradiation hardening was due to thedislocation loops.

In the above equation, οߪ௖௔௟௖ is the increase in yieldstress (MPa), ‫ ܯ‬is Taylor factor (3.06), ߙ is strength
factor of dislocation loops (0.5) [9], ߤ is shear modulus(76 GPa), ܾ is Burgers vector (2.5 ൈ 10-8 cm), ܰ isnumber density of dislocation loops, ݀ is size ofdislocation loops. The increase in Vickers hardness ο‫ܪ‬௏is converted into that of yield stress using the following
equation [10]:86Y. Sakai et al., Effects of Ta and C on Irradiation Hardening of Austenitic Stainless SteelFig. 9 The images of TEM irradiated at 473K to 673K.

4. Summary

3.4 Effect of Ta and C
The contribution of Ta and C to the hardening ofspecimens irradiated to 0.3 dpa at 573K is shown in Fig.
13. Tantalum suppresses the irradiation hardening ofstainless steels as shown in Fig. 7 and Fig. 8.
Irradiation hardening increased with increasing C
concentration. It is reported that dislocation loop density
increased with increasing C concentration [11]. Thismeans that the increase in irradiation hardening with C
concentration observed in this study would be due to theincrease in dislocation loop density. However, theirradiation hardening of the alloys is almost the same asthat of 304L.
The effect of the difference of Ta/C ratio on theirradiation hardening was not observed in Fig.13.

In order to investigate the effect of Ta and C on
irradiation hardening of stainless steels underdevelopment, Vickers hardness test and TEM observation
were carried out after 1 MeV H+ ion irradiation. The
following results were obtained.

1.

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Irradiation hardening was observed in all alloys.
The irradiation hardening of the alloys underdevelopment was almost the same as that of304L, and the effect of Ta and C addition on
irradiation hardening was not observed clearly
within the investigated composition range. So it
is expected that Ta and C addition alloy
increase IASCC resistance without affecting

Y. Sakai et al., Effects of Ta and C on Irradiation Hardening of Austenitic Stainless Steel

References

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Nuclear Materials, 189, 167-174 (1992)[5] S. Matsuyama et al. , Nuclear Instruments and Methods inPhysics Research B, 267, 2060-2064 (2009)[6] http://www.srim.org[7] K. Yabuuchi et al., Journal of Nuclear Materials, 446,142-147 (2014)[8] E. Orowan, Symposium on Internal Stress in Metals
Alloys, 451, 1948
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The irradiation hardening was explained bydislocation loop formation.

Acknowledgement

The authors appreciate prof. Matsuyama, Mr.
Fujisawa and Mr. Ishiya of Tohoku University forsupporting ion irradiation experiments.

Fig. 10 Average dislocation loop size vs.
irradiation temperature.

Fig. 11 Dislocation loop density vs. irradiation
temperature.

Fig. 12 Relationship between the yield stresscalculated from microstructure andmeasured by Vickers hardness.

Fig. 13 Irradiation hardening as a function of Ta andC concentration.

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