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Evaluation of irradiation assisted stress corrosion cracking (IASCC) of type 316 stainless steel irradiated in FBR
journal of nuclear materials
Journal of Nuclear Materials 207 (1993) 159-168 North-Holland
Evaluation of irradiation assisted stress corrosion cracking (IASCC) of type 316 stainless steel irradiated in FBR T. Tsukada a, S. Jitsukawa a, K. Shiba a, Y. Sato b, I. Shibahara b and H. Nakajima a aJapan Atomic Energy Research Institute, Tokai-mura, Ibaraki-ken 319-11, Japan b Power Reactor and Nuclear Fuel Development Corporation, Oarai-machi, Ibaraki-ken 311-13, Japan
Received 18 February 1993; accepted 5 July 1993
Type 316 stainless steel from the core of the experimental fast breeder reactor (FBR) JOY0 was examined by the slow strain rate tensile (SSRT) test in pure, oxygenated-water and air and by the electrochemical potentiokinetic reactivation CEPR) test to evaluate a susceptibility to the irradiation assisted stress corrosion cracking (IASCC) and the radiation-induced segregation (RIS). The solution annealed and 20% cold-worked materials had been irradiated at 425°C to a neutron fluence of 8.3 X 10z6 n/m* ( > 0.1 MeVl which is equivalent to 40 displacement per atom (dpa). Intergranular cracking was induced by the SSRT in water at 200 and 300°C but was not observed on specimen tested in water at 60°C and in air at 300°C. This indicates that irradiation increased a susceptibility to stress corrosion cracking (SCC) in water. After the EPR test, grain boundary etching was observed in addition to grain face etching. This suggests Cr depletion may have occurred both at grain boundary and at defect clusters during the irradiation. The results are compared with the behavior of similar materials irradiated with different neutron spectrum.
1. Introduction assisted stress corrosion Irradiation cracking (IASCC) has been considered as a degradation phe-
nomenon for the core internals of light water reactor (LWR) [l-2]. Slow strain rate tensile (SSRTI tests in high temperature water on the materials irradiated in LWR have been, therefore, performed extensively to evaluate susceptibilities to IASCC [3-41. It has been suggested that radiation-induced segregation (RIS) of major alloying elements such as Cr and minor impurity elements such as P and Si may play an important role in the mechanism of IASCC. However the experimental results were sometimes contrary to each other on this point and confuse discussions regarding IASCC. The authors have extended the study of IASCC to neutron spectra different from that of LWR. Radiation conditions are obviously a significant factor but information from outside LWR conditions has not been previously available. Therefore, we have previously performed SSRT tests on the material irradiated under the spectrally-tailored conditions that simulate some aspects of the irradiation environment expected on the first-wall structure of a fusion reactor [5]. In the pre0022-3115/93/$06.00
sent work, fuel assembly duct material irradiated in fast breeder reactor (FBR) was examined by SSRT and electrochemical potentiokinetic reactivation CEPR) tests. The former was carried out to evaluate IASCC susceptibility and the latter was performed to investigate RIS. Basic process of radiation-induced damage consists of atomic displacements and transmutation reactions. A reliable parameter to characterize neutron irradiation condition is a ratio of He (at. ppm) generated due to transmutation and damage accumulation (displacement per atom, dpa). Although there are various irradiation factors affecting IASCC, an approach from the aspect of neutron spectrum in terms of the He/dpa ratio may offer important information for understanding the IASCC phenomenon.
2. Experimental 2.1. Material and irradiation condition The duct of fuel assembly for FBR JOY0 is in the shape of hexagonal tube and made of type 316 stainless
0 1993 - Elsevier Science Publishers B.V. All rights reserved
160
T. Tsukada et al. / OiSCC of type 316 SS irradiated in FBR
steel with a thickness of 1.9 mm. Chemical composition of the duct material is shown in table 1. Since the material was manufactured by a double vacuum melting process, the amounts of the gas elements, i.e. 0 and N, were reduced to 8 mass ppm 0 and 80 mass ppm N. Solution annealed treatment at 1090°C for 25 min and 20% cold working were performed at manufacturing process. The material was irradiated as a component of driver fuel assembly in JOY0 for 17 cycles (421 fullpower-days) at 100 MW. After the removal from the core, the fuel assembly was stored in the inert gas environment for about two years before post irradiation examination (PIE). Neutron fluence of the examined material was 8.3 x 10z6 n/m2 (> 0.1 MeV) which is equivalent to 40 dpa. Irradiation temperature was 425 + 15°C and the range of temperature is due to temperature gradient in the flowing liquid sodium. The amount of He in the material was estimated to be about 15 at.ppm. 2.2. SSRT test A configuration of the specimen for SSRT test is shown in fig. 1. Specimens were stamped out by 20 ton press machine and cleaned with acetone in ultrasonic bath before tests. The SSRT machine with a high temperature and high pressure water supply system installed at the JAERI hot laboratory was used [6]. The tests were conducted at 60, 200 and 300°C in the high purity water at 9.3 MPa and also at 300°C in air. Dissolved oxygen content in the water was controlled at 32 mass ppm and the electric conductivity was below 0.2 kS/cm at inlet and below 1 t&/cm at outlet of the autoclave. Flow rate of water was 5 l/h. The extension rate was controlled at 1.53 x 10e4 mm/min which translates to a strain rate of 1.7 x 10m7 s-l on the specimen. Specimen was electrically insulated from the test machine with Zircaloy adapters oxidized on the surface. A load cell is located outside autoclave and, therefore, a friction at the pressure compensation mechanism was included on the measured load. The friction can be estimated as a remaining load just after
Fig. 1. SSRT test specimen (mm).
a fracture of specimen and was in a range of 0.3 to 0.4 kN. It was subtracted from the load measured throughout the test to obtain a stress on the specimen more accurately. The fractured surface of the specimen was examined using both optical and scanning electron microscopes (SEMI. 2.3. Electrochemical potentiokinetic test
reactiuation CEPR)
Test apparatus for the electrochemical measurement was installed at JAERI hot laboratory and EPR tests were carried out by remote operation in the hot cell. Details of the apparatus were reported elsewhere [6-71. Irradiated material cut into 10 X 10 mm was mounted in epoxy resin. The specimen was polished finally with diamond paste. It is known that in case of the irradiated specimen grain face etching is the main cause of reactivation peak current and a contribution of grain boundary etching becomes relatively small [S]. Therefore, to detect a grain boundary sensitization more clearly a composition of the test solution was modified from that of the standard single loop EPR test method [9]. The concentration of KSCN in 0.5 kmol/m3 H,SO, was increased to 0.1 kmol/m3 from 0.01 kmol/m3 of the standard EPR solution. This modification had been applied by Iwabuchi et al. [lo] to detect a narrow Cr depletion band at grain boundary after thermal sensitization simulating RIS. Temperature of test solution was controlled at 30°C and the scan rate of electric potential was 100 mV/min. Before scanning of potential the specimen was passivated at +200 mV versus saturated calomel electrode (SCE)
Table 1 Chemical composition of the specimen material of solution annealed and 20% cold-worked (CW) type 316 stainless steel for the duct of FBR JOY0 (mass%) C
Si
Mn
P
s
Ni
Cr
MO
Ti
B
Fe
0.053
0.72
1.97
0.02
0.004
13.46
16.35
2.49
0.068
0.0022
bal.
161
T. Tsukada et al. / IASCC of type316 SS irradiatedin FBR
for 2 min. After EPR tests, the surface of specimen was examined using SEM.
3. Results
3.1. Stress-strain curves Characteristic values and engineering stress-strain curves obtained on the four irradiated specimens are summarized in table 2 and fig. 2. To calculate uniform elongation, an elastic part was subtracted from the strain at ultimate strength, but it was not subtracted to calculate a total elongation because a fracture load was not obvious on the curves. The elongations obtained from the tests in the water decreased with increasing test temperatures. Specimen tested at 60°C showed the largest total elongation of 15.2% and ultimate strength of 942 MPa. Specimen tested at 200°C in water and at 300°C in air showed comparable total elongations of about 9%. On the curve of the specimen tested at 2Oo”C, a few stress drops were recorded. They appeared to be caused by the environmental assisted cracking. Specimen tested at 300°C in water showed the smallest elongation and nearly no uniform elongation. The difference between the curves from specimens tested at 300°C in water and air may imply an environmental effect due to the presence of high temperature water. 3.2. Fracture surfaces Fig. 3 shows the entire fracture surface of the specimen tested at 60°C in water. The specimen was
Engineering strain (Or,)
Fig. 2. Stress-strain
curves from the irradiated FBR duct material tested in oxygenated high purity water (W) and in air (A).
fully fractured in ductile mode and dimple patterns were observed on the fracture surface. Reduction in width was observed near fractured portion and it corresponded with the large elongation, On the specimen tested at 200°C in water, approximately 5% of surface area exhibited cracking by intergranular and transgranular mixed mode, as shown in figs. 4(a) and 4(b). The other portions of the fracture surface showed ductile fracture and the fracture surface slanted about 45” from the tensile direction as seen in fig. 5 that shows a side of the specimen. In fig. 5, a few small cracks were observed and appear to be corresponding to the load drops on the stress-strain curve. Fig. 6(a) shows the mixed mode of,intergranular and transgranular cracking and fig. 6(b) shows the ductile area.
Table 2 Summary of SSRT test results for FBR duct material irradiated at 425°C to 8.3X 1O26n/m* (E > 0.1 MeV) that is equivalent to 40 dpa Test environment/ temp. CC)
0.2% Proof stress (MPa)
Ultimate stress (MPa)
Uniform elongation (%o)
Total elongation (%)
Fracture mode a
lG+TG cracking fraction (%)
Water/60 Water/200 Water/300 Air/300
810 833 763 710
942 942 787 828
6.0 3.3 0.4 4.2
15.2 9.1 5.0 9.6
D IG+TG+D IG+D D
0 approx. 5 approx. 25 0
a D: ductile; IG: intergranular, TG: transgranular.
T Tsukada et al. / IASCC of lype 316 SS irradiated in FBR
162
Fig. 3. Ductile
Fig. 4. SEM photographs
fracturd
of (a) fractured
surface
of the specimen
after SSRT test carried
out at 60°C in water.
surface
of the specimen tested at 200°C in water and (h) the area failed by intergranular and transgranular cracking.
T. Tsukada et al. / USCC
Fig. 5. Optical microphotograph
of type 316 SS irradiated in FBR
163
of the side of specimen tested at 200°C in water. Sheared fracture and cracks indicated by arrows are noted.
The intergranular cracking was also observed on the specimen tested at 300°C in water. As shown in fig.
7(a), two parts near the edge were cracked by intergranular mode. The total area fraction of the intergranular cracking was roughly 25%. No transgranular cracking was observed on this specimen. Fig. 7(b) shows a part of intergranular cracking. The shape of fractured surface is irregular as shown in fig. 8 and the feature is similar to that observed on the specimen tested at 200°C. Intergranular cracking area is magnified in fig. 9(a). Fig. 9(b) shows the ductile fracture surface where deformed dimple-like pattern was observed Fig. 10 shows the ductile fracture surface of the specimen tested at 300°C in air. No intergranular cracking was observed in this environment. This confirms that the intergranular cracking occurred in water was an environmental effect of high temperature water.
3.3. EPR tests Fig. 11 shows the polarization reactivation curves obtained from the standard and modified EPR tests. Reactivation peaks due to sensitization in the alloy were observed on both curves. The modified EPR test showed a higher peak current density. In fig. 12, SEM photographs of the specimen surfaces after EPR tests are shown. Nonuniform etching of grain faces and slight etching of grain boundaries were observed after the standard EPR test as shown in figs. 12(a) and 12(b). On the other hand, grain faces were etched uniformly by the modified method as seen in fig. 12(c) and grain boundary etching was observed distinctly in
Fig. 6. Magnified views of (a) intergranular and transgranular cracking and (b) ductile fractured area on the specimen tested at 200°C in water.
T. Tsukada et al. / MSCC of type 316 SS irradiated in FBR
164
Fig. 7. SEM
photographs
of (a) fractured
surface
of the specimen tested intergranular cracking.
at 300°C
in the water
and
(b) the area
failed
by
T. Tsukada et al. / IASCC of type 316 SS irradiated in FBR
Fig. 9. Magnified views of (a) intergranular cracking and (b) ductile fractured area on the specimen tested at 300°C in water.
fig. 12(d). The width of the grain boundary etching was estimated to be approximately 0.2 to 0.5 km.
4. Discussion The occurrence of stress corrosion cracking (SCC) was confirmed by a comparison of fracture surfaces of the specimens tested at 300°C both in water and in air. Intergranular cracks were observed on the surface of specimens tested in the water. Small cracks were formed in the gage region of the specimens tested in water at 200 and 300°C however, few cracks were
165
observed on the specimen tested at 60°C. Since most of the planes of these small cracks were perpendicular to the tensile direction, the tensile straining affected these cracks and it is suggested that these cracks were formed by SCC. Observations of the fracture surfaces after tests in high temperature water showed that the specimens failed by both grain boundary separation and ductile shearing. Sizes of the intergranular cracks were slightly larger than, but not very different from, those of the small cracks formed over the gage region. It appears that fracture of the specimen occurred by shearing between two cracked portions when the stress in the area between the cracks increased to cause plastic shearing. Therefore, an area1 fraction of intergranular cracking in the fracture surface may correlate with a susceptibility to SCC of the irradiated material. In the separate study [5], a solution annealed type 316 stainless steel had been irradiated to 8 dpa in the Oak Ridge Research Reactor (ORR) under the spectrally-tailored condition and examined by SSRT tests in oxygenated high purity water. The test conditions were the same with those of the present study. A specimen irradiated at 400°C in ORR and tested at 300°C showed a fracture by 100% intergranular cracking mode. The largest fraction of intergranular cracking obtained in this study was 25% on the specimen tested at 300°C in water and it was smaller than that found on the specimen irradiated in ORR. This indicates that ORR irradiation to 8 dpa caused larger IASCC susceptibility than JOY0 irradiation to 40 dpa. Differences of neutron spectra and material condition (solution annealed versus cold-worked) may have caused this inverse dependence on displacement damage. Narrow etched ditches at grain boundaries and etch pits on grain faces were formed during EPR tests on the surfaces of specimens. Similar ditches after EPR test were found for the specimens irradiated in ORR at 400°C by Inazumi et al. [ll]. They also carried out EPR test and analytical electron microscopy (AEM) on the material irradiated at 420°C to 9 dpa in the Fast Flux Test Facility (FFTF) [8,11-121. On that material, the reactivation by EPR test was observed and Cr depletion at grain boundaries, voids and faulted dislocation loops were detected by AEM. They concluded that the decrease in corrosion resistance at grain boundaries and grain faces was caused by RIS. Ditches at grain boundaries found in the present study may also be the result of segregation at grain boundaries during irradiation. The narrow width of etching at grain boundary might be a characteristic of Cr depletion by RIS because a width of Cr depletion at grain
166
T. Tsukada et al. / IASCC of type 316 SS irradiated in FBR
Fig. 10. Ductile fracture surface of the specimen aft& SSRT test carried out at 300°C in air. boundary on the neutron irradiated specimen was reported in the range of a few nanometer [12]. Dislocations are known to absorb radiation-produced interstitials preferentially during irradiation. This causes vacancies to flow to neutral sinks, such as cavities and grain boundaries. Vacancy flux to cavities causes cavity growth that leads to swelling of alloy and Ni-enrichment and Cr-depletion at cavity surfaces. The flow of vacancies to grain boundaries may also cause the same type of segregation during irradiation. Neutron spectrum of ORR was adjusted to achieve He/dpa ratio relevant to the fusion reactor first wall for Nibearing austenitic stainless steels [13]. The He/dpa
Potential (mV vs. SCE) Fig. 11. Polarization reactivation curves for FBR duct material obtained from EPR tests in standard and modified test solutions.
ratio of about 12.5 for ORR is larger than that of about 0.4 for JOY0 irradiation. Stoller et al. [14] indicated that vacancy flow to neutral sinks becomes maximum at a He/dpa ratio close to that expected in the fusion neutron environment. Sawai et al. [15] also reported that relatively large swelling of 0.3% occurred by ORR spectrally-tailored irradiation to about 8 dpa even at a low temperature of 400°C. Larger vacancy flux during ORR spectrally-tailored irradiation may cause larger swelling and more pronounced enrichment of Ni and depletion of Cr at grain boundaries. Swelling found in the fuel pin cladding irradiated in a similar condition with the duct material in this study was less than 0.1%. This suggests that segregation at grain boundaries in the duct material was smaller than that in the material irradiated in ORR. The duct material for JOY0 was cold worked before irradiation. Cold working introduces a large dislocation density. Although dislocations act preferential sinks for interstitial, strong sink strength of high density dislocations introduced by cold working reduces the vacancy concentration during irradiation and suppresses swelling. Therefore, cold working may also suppress the segregation and reduce IASCC susceptibility. This may also explain the dependence in IASCC susceptibility between the ORR and JOY0 irradiated materials. Under the LWR operation conditions, on the other hand, swelling does not occur because of lower fast neutron fluence and temperature. However, a susceptibility to IASCC of the material irradiated in LWR seems to be higher than that of the materials irradiated
T. Tsukada et al. / IASCC of type 316 SS irradiated in FBR
in ORR and JOY0 [3,4]. Further extension of these results to the LWR irradiation condition is difficult at present blut probably other effects, e.g. impurity segrerole for LWR gation, nnay also play an important materials
167
5. Conclusions Type 316 stainless steel from the core of tlle exl lerimental FBR “JOYO” was examined by the 5;SRT test in pure, oxygenated water and air and by the EPR test
Fig. 12. Specimen surfaces after EPR tests observed by SEM. (a) and (b) show nonuniform etching of grain faces and slight etching of grain boundaries resulted from standard EPR test. Cc) and (d) show uniform etching and narrow ditches formed at grain boundaries after modified EPR test.
168
i? Tsukada et al. / L4SCC
for the study of IASCC. The material was irradiated at 425°C to a neutron fluence of 8.3 x 102” n/m2 (> 0.1 MeV) which is equivalent to 40 dpa. The results are summarized as follows: (1) Intergranular cracking was observed on the fractured specimen tested by SSRT at 200 and 300°C in water. Area1 fractions of intergranular cracking portions in the fracture surface were increased with increasing test temperatures that were 0, 5 and 25% at 60, 200 and 3OO”C,respectively. (2) The specimen tested by SSRT at 300°C in air showed fully ductile fracture and, therefore, the intergranular cracking observed in water is considered as an evidence of IASCC. (3) Fracture of the specimens in high temperature water occurred by shearing between intergranular cracking portions caused by SCC. On the sheared surface, ductile-dimple pattern was observed. (4) After the EPR test, grain boundary etching was observed in addition to grain face etching. This may suggest Cr-depletions at grain boundaries and defect clusters. (5) Comparison with the SSRT test results from the spectrally-tailored irradiated type 316 material suggested a similar dependency between a susceptibility to IASCC and a swelling on He/dpa ratio. Differences of neutron spectrum and materials condition are considered to cause these dependencies.
Acknowledgements
The authors wish to acknowledge Dr. A. Hishinuma of JAERI for his helpful support and very grateful to staff members of the hot laboratories of JAERI and PNC. We are very much indebted to Prof. T. Shoji of Tohoku University for his valuable suggestions on the electrochemical test techniques and to Dr. G.E.C. Bell of Oak Ridge National Laboratory (ORNL) for his constructive advice. We also thank Drs. T. Kondo of JAERI, M. Katsuragawa of PNC and K. Tomabechi (formerly staff of JAERI) for their encouragement.
of type 316 SS irradiated in FBR References [l] J.L. Nelson and P.L. Andresen, Proc. 5th Int. Symp. on Environmental Degradation of Materials in Nuclear Power Systems - Water Reactors CANS, Monterey, California, 1991) p. 10. [2] H. Hanninen and I. Aho-Mantila, Proc. 3rd Int. Symp. on Environmental Degradation of Materials in Nuclear Power Systems - Water Reactors (AIME, Traverse City, Michigan, 1987) p. 77. [3] A.J. Jacobs, O.P. Wozadlo, K. Nakata, T. Yoshida and I. Masaoka, ibid. ref. [2], p. 673. [4] M. Kodama, S. Nishimura, J. Morisawa, S. Shima, S. Suzuki and M. Yamamoto, Proc. 5th Int. Symp. on Environmental Degradation of Materials in Nuclear Power Systems - Water Reactors CANS, Monterey, California, 19911 p. 948. [5] T. Tsukada, K. Shiba, G.E.C. Bell and H. Nakajima, Corrosion/92 (NACE, Nashville, TN, 1992) paper no. 104. [6] T. Tsukada, K. Shiba, M. Ohmi, M. Kizaki, H. Matsushima and H. Nakajima, Proc. 3rd Asian Symp. on Research Reactor (JAERI, Hitachi, Japan, 1991) p. 621. [7] K. Shiba, T. Tsukada, H. Nakajima, H. Matsushima, I. Takahashi, K. Sonobe and T. Komatsu, Japan Atomic Energy Research Institute Report, JAERI-M 91-024 (19911, in Japanese. [S] T. Inazumi, G.E.C. Bell, E.A. Kenik and K. Kiuchi, Corrosion 46 (10) (1990) 786. [9] ASTM standard G 108-92 (ASTM, 1992). [lo] Y. Iwabuchi, T. Fujimoto, Y. Watanabe and T. Shoji, Zairyo-to-Kankyo (Corrosion Engineering) 42 (1993) 2, in Japanese. [ll] T. Inazumi, G.E.C. Bell, P.J. Maziasz and T. Kondo, J. Nucl. Mater. 191-194 (1992) 1018. [12] E.A. Kenik, T. Inazumi and G.E.C. Bell, J. Nucl. Mater. 183 (1991) 145. [13] M.L. Grossbeck, E.E. Bloom, J.W. Woods, J.M. Vitek and K.R. Thorns, Proc. Conf. on Fast, Thermal and Fusion Reactor Experiments, vol. 1 CANS, Salt Lake City, Utah, 1982) pp. 1-199. [14] R.E. StoIIer, P.J. Maziasz and A.F. Rowcliffe, J. Nucl. Mater. 155-157 (1988) 1328. [15] T. Sawai, P.J. Maziasz, H. Kanazawa and A. Hishinuma, J. Nucl. Mater. 191-194 (1992) 712.
Journal of Nuclear Materials 207 (1993) 159-168 North-Holland
Evaluation of irradiation assisted stress corrosion cracking (IASCC) of type 316 stainless steel irradiated in FBR T. Tsukada a, S. Jitsukawa a, K. Shiba a, Y. Sato b, I. Shibahara b and H. Nakajima a aJapan Atomic Energy Research Institute, Tokai-mura, Ibaraki-ken 319-11, Japan b Power Reactor and Nuclear Fuel Development Corporation, Oarai-machi, Ibaraki-ken 311-13, Japan
Received 18 February 1993; accepted 5 July 1993
Type 316 stainless steel from the core of the experimental fast breeder reactor (FBR) JOY0 was examined by the slow strain rate tensile (SSRT) test in pure, oxygenated-water and air and by the electrochemical potentiokinetic reactivation CEPR) test to evaluate a susceptibility to the irradiation assisted stress corrosion cracking (IASCC) and the radiation-induced segregation (RIS). The solution annealed and 20% cold-worked materials had been irradiated at 425°C to a neutron fluence of 8.3 X 10z6 n/m* ( > 0.1 MeVl which is equivalent to 40 displacement per atom (dpa). Intergranular cracking was induced by the SSRT in water at 200 and 300°C but was not observed on specimen tested in water at 60°C and in air at 300°C. This indicates that irradiation increased a susceptibility to stress corrosion cracking (SCC) in water. After the EPR test, grain boundary etching was observed in addition to grain face etching. This suggests Cr depletion may have occurred both at grain boundary and at defect clusters during the irradiation. The results are compared with the behavior of similar materials irradiated with different neutron spectrum.
1. Introduction assisted stress corrosion Irradiation cracking (IASCC) has been considered as a degradation phe-
nomenon for the core internals of light water reactor (LWR) [l-2]. Slow strain rate tensile (SSRTI tests in high temperature water on the materials irradiated in LWR have been, therefore, performed extensively to evaluate susceptibilities to IASCC [3-41. It has been suggested that radiation-induced segregation (RIS) of major alloying elements such as Cr and minor impurity elements such as P and Si may play an important role in the mechanism of IASCC. However the experimental results were sometimes contrary to each other on this point and confuse discussions regarding IASCC. The authors have extended the study of IASCC to neutron spectra different from that of LWR. Radiation conditions are obviously a significant factor but information from outside LWR conditions has not been previously available. Therefore, we have previously performed SSRT tests on the material irradiated under the spectrally-tailored conditions that simulate some aspects of the irradiation environment expected on the first-wall structure of a fusion reactor [5]. In the pre0022-3115/93/$06.00
sent work, fuel assembly duct material irradiated in fast breeder reactor (FBR) was examined by SSRT and electrochemical potentiokinetic reactivation CEPR) tests. The former was carried out to evaluate IASCC susceptibility and the latter was performed to investigate RIS. Basic process of radiation-induced damage consists of atomic displacements and transmutation reactions. A reliable parameter to characterize neutron irradiation condition is a ratio of He (at. ppm) generated due to transmutation and damage accumulation (displacement per atom, dpa). Although there are various irradiation factors affecting IASCC, an approach from the aspect of neutron spectrum in terms of the He/dpa ratio may offer important information for understanding the IASCC phenomenon.
2. Experimental 2.1. Material and irradiation condition The duct of fuel assembly for FBR JOY0 is in the shape of hexagonal tube and made of type 316 stainless
0 1993 - Elsevier Science Publishers B.V. All rights reserved
160
T. Tsukada et al. / OiSCC of type 316 SS irradiated in FBR
steel with a thickness of 1.9 mm. Chemical composition of the duct material is shown in table 1. Since the material was manufactured by a double vacuum melting process, the amounts of the gas elements, i.e. 0 and N, were reduced to 8 mass ppm 0 and 80 mass ppm N. Solution annealed treatment at 1090°C for 25 min and 20% cold working were performed at manufacturing process. The material was irradiated as a component of driver fuel assembly in JOY0 for 17 cycles (421 fullpower-days) at 100 MW. After the removal from the core, the fuel assembly was stored in the inert gas environment for about two years before post irradiation examination (PIE). Neutron fluence of the examined material was 8.3 x 10z6 n/m2 (> 0.1 MeV) which is equivalent to 40 dpa. Irradiation temperature was 425 + 15°C and the range of temperature is due to temperature gradient in the flowing liquid sodium. The amount of He in the material was estimated to be about 15 at.ppm. 2.2. SSRT test A configuration of the specimen for SSRT test is shown in fig. 1. Specimens were stamped out by 20 ton press machine and cleaned with acetone in ultrasonic bath before tests. The SSRT machine with a high temperature and high pressure water supply system installed at the JAERI hot laboratory was used [6]. The tests were conducted at 60, 200 and 300°C in the high purity water at 9.3 MPa and also at 300°C in air. Dissolved oxygen content in the water was controlled at 32 mass ppm and the electric conductivity was below 0.2 kS/cm at inlet and below 1 t&/cm at outlet of the autoclave. Flow rate of water was 5 l/h. The extension rate was controlled at 1.53 x 10e4 mm/min which translates to a strain rate of 1.7 x 10m7 s-l on the specimen. Specimen was electrically insulated from the test machine with Zircaloy adapters oxidized on the surface. A load cell is located outside autoclave and, therefore, a friction at the pressure compensation mechanism was included on the measured load. The friction can be estimated as a remaining load just after
Fig. 1. SSRT test specimen (mm).
a fracture of specimen and was in a range of 0.3 to 0.4 kN. It was subtracted from the load measured throughout the test to obtain a stress on the specimen more accurately. The fractured surface of the specimen was examined using both optical and scanning electron microscopes (SEMI. 2.3. Electrochemical potentiokinetic test
reactiuation CEPR)
Test apparatus for the electrochemical measurement was installed at JAERI hot laboratory and EPR tests were carried out by remote operation in the hot cell. Details of the apparatus were reported elsewhere [6-71. Irradiated material cut into 10 X 10 mm was mounted in epoxy resin. The specimen was polished finally with diamond paste. It is known that in case of the irradiated specimen grain face etching is the main cause of reactivation peak current and a contribution of grain boundary etching becomes relatively small [S]. Therefore, to detect a grain boundary sensitization more clearly a composition of the test solution was modified from that of the standard single loop EPR test method [9]. The concentration of KSCN in 0.5 kmol/m3 H,SO, was increased to 0.1 kmol/m3 from 0.01 kmol/m3 of the standard EPR solution. This modification had been applied by Iwabuchi et al. [lo] to detect a narrow Cr depletion band at grain boundary after thermal sensitization simulating RIS. Temperature of test solution was controlled at 30°C and the scan rate of electric potential was 100 mV/min. Before scanning of potential the specimen was passivated at +200 mV versus saturated calomel electrode (SCE)
Table 1 Chemical composition of the specimen material of solution annealed and 20% cold-worked (CW) type 316 stainless steel for the duct of FBR JOY0 (mass%) C
Si
Mn
P
s
Ni
Cr
MO
Ti
B
Fe
0.053
0.72
1.97
0.02
0.004
13.46
16.35
2.49
0.068
0.0022
bal.
161
T. Tsukada et al. / IASCC of type316 SS irradiatedin FBR
for 2 min. After EPR tests, the surface of specimen was examined using SEM.
3. Results
3.1. Stress-strain curves Characteristic values and engineering stress-strain curves obtained on the four irradiated specimens are summarized in table 2 and fig. 2. To calculate uniform elongation, an elastic part was subtracted from the strain at ultimate strength, but it was not subtracted to calculate a total elongation because a fracture load was not obvious on the curves. The elongations obtained from the tests in the water decreased with increasing test temperatures. Specimen tested at 60°C showed the largest total elongation of 15.2% and ultimate strength of 942 MPa. Specimen tested at 200°C in water and at 300°C in air showed comparable total elongations of about 9%. On the curve of the specimen tested at 2Oo”C, a few stress drops were recorded. They appeared to be caused by the environmental assisted cracking. Specimen tested at 300°C in water showed the smallest elongation and nearly no uniform elongation. The difference between the curves from specimens tested at 300°C in water and air may imply an environmental effect due to the presence of high temperature water. 3.2. Fracture surfaces Fig. 3 shows the entire fracture surface of the specimen tested at 60°C in water. The specimen was
Engineering strain (Or,)
Fig. 2. Stress-strain
curves from the irradiated FBR duct material tested in oxygenated high purity water (W) and in air (A).
fully fractured in ductile mode and dimple patterns were observed on the fracture surface. Reduction in width was observed near fractured portion and it corresponded with the large elongation, On the specimen tested at 200°C in water, approximately 5% of surface area exhibited cracking by intergranular and transgranular mixed mode, as shown in figs. 4(a) and 4(b). The other portions of the fracture surface showed ductile fracture and the fracture surface slanted about 45” from the tensile direction as seen in fig. 5 that shows a side of the specimen. In fig. 5, a few small cracks were observed and appear to be corresponding to the load drops on the stress-strain curve. Fig. 6(a) shows the mixed mode of,intergranular and transgranular cracking and fig. 6(b) shows the ductile area.
Table 2 Summary of SSRT test results for FBR duct material irradiated at 425°C to 8.3X 1O26n/m* (E > 0.1 MeV) that is equivalent to 40 dpa Test environment/ temp. CC)
0.2% Proof stress (MPa)
Ultimate stress (MPa)
Uniform elongation (%o)
Total elongation (%)
Fracture mode a
lG+TG cracking fraction (%)
Water/60 Water/200 Water/300 Air/300
810 833 763 710
942 942 787 828
6.0 3.3 0.4 4.2
15.2 9.1 5.0 9.6
D IG+TG+D IG+D D
0 approx. 5 approx. 25 0
a D: ductile; IG: intergranular, TG: transgranular.
T Tsukada et al. / IASCC of lype 316 SS irradiated in FBR
162
Fig. 3. Ductile
Fig. 4. SEM photographs
fracturd
of (a) fractured
surface
of the specimen
after SSRT test carried
out at 60°C in water.
surface
of the specimen tested at 200°C in water and (h) the area failed by intergranular and transgranular cracking.
T. Tsukada et al. / USCC
Fig. 5. Optical microphotograph
of type 316 SS irradiated in FBR
163
of the side of specimen tested at 200°C in water. Sheared fracture and cracks indicated by arrows are noted.
The intergranular cracking was also observed on the specimen tested at 300°C in water. As shown in fig.
7(a), two parts near the edge were cracked by intergranular mode. The total area fraction of the intergranular cracking was roughly 25%. No transgranular cracking was observed on this specimen. Fig. 7(b) shows a part of intergranular cracking. The shape of fractured surface is irregular as shown in fig. 8 and the feature is similar to that observed on the specimen tested at 200°C. Intergranular cracking area is magnified in fig. 9(a). Fig. 9(b) shows the ductile fracture surface where deformed dimple-like pattern was observed Fig. 10 shows the ductile fracture surface of the specimen tested at 300°C in air. No intergranular cracking was observed in this environment. This confirms that the intergranular cracking occurred in water was an environmental effect of high temperature water.
3.3. EPR tests Fig. 11 shows the polarization reactivation curves obtained from the standard and modified EPR tests. Reactivation peaks due to sensitization in the alloy were observed on both curves. The modified EPR test showed a higher peak current density. In fig. 12, SEM photographs of the specimen surfaces after EPR tests are shown. Nonuniform etching of grain faces and slight etching of grain boundaries were observed after the standard EPR test as shown in figs. 12(a) and 12(b). On the other hand, grain faces were etched uniformly by the modified method as seen in fig. 12(c) and grain boundary etching was observed distinctly in
Fig. 6. Magnified views of (a) intergranular and transgranular cracking and (b) ductile fractured area on the specimen tested at 200°C in water.
T. Tsukada et al. / MSCC of type 316 SS irradiated in FBR
164
Fig. 7. SEM
photographs
of (a) fractured
surface
of the specimen tested intergranular cracking.
at 300°C
in the water
and
(b) the area
failed
by
T. Tsukada et al. / IASCC of type 316 SS irradiated in FBR
Fig. 9. Magnified views of (a) intergranular cracking and (b) ductile fractured area on the specimen tested at 300°C in water.
fig. 12(d). The width of the grain boundary etching was estimated to be approximately 0.2 to 0.5 km.
4. Discussion The occurrence of stress corrosion cracking (SCC) was confirmed by a comparison of fracture surfaces of the specimens tested at 300°C both in water and in air. Intergranular cracks were observed on the surface of specimens tested in the water. Small cracks were formed in the gage region of the specimens tested in water at 200 and 300°C however, few cracks were
165
observed on the specimen tested at 60°C. Since most of the planes of these small cracks were perpendicular to the tensile direction, the tensile straining affected these cracks and it is suggested that these cracks were formed by SCC. Observations of the fracture surfaces after tests in high temperature water showed that the specimens failed by both grain boundary separation and ductile shearing. Sizes of the intergranular cracks were slightly larger than, but not very different from, those of the small cracks formed over the gage region. It appears that fracture of the specimen occurred by shearing between two cracked portions when the stress in the area between the cracks increased to cause plastic shearing. Therefore, an area1 fraction of intergranular cracking in the fracture surface may correlate with a susceptibility to SCC of the irradiated material. In the separate study [5], a solution annealed type 316 stainless steel had been irradiated to 8 dpa in the Oak Ridge Research Reactor (ORR) under the spectrally-tailored condition and examined by SSRT tests in oxygenated high purity water. The test conditions were the same with those of the present study. A specimen irradiated at 400°C in ORR and tested at 300°C showed a fracture by 100% intergranular cracking mode. The largest fraction of intergranular cracking obtained in this study was 25% on the specimen tested at 300°C in water and it was smaller than that found on the specimen irradiated in ORR. This indicates that ORR irradiation to 8 dpa caused larger IASCC susceptibility than JOY0 irradiation to 40 dpa. Differences of neutron spectra and material condition (solution annealed versus cold-worked) may have caused this inverse dependence on displacement damage. Narrow etched ditches at grain boundaries and etch pits on grain faces were formed during EPR tests on the surfaces of specimens. Similar ditches after EPR test were found for the specimens irradiated in ORR at 400°C by Inazumi et al. [ll]. They also carried out EPR test and analytical electron microscopy (AEM) on the material irradiated at 420°C to 9 dpa in the Fast Flux Test Facility (FFTF) [8,11-121. On that material, the reactivation by EPR test was observed and Cr depletion at grain boundaries, voids and faulted dislocation loops were detected by AEM. They concluded that the decrease in corrosion resistance at grain boundaries and grain faces was caused by RIS. Ditches at grain boundaries found in the present study may also be the result of segregation at grain boundaries during irradiation. The narrow width of etching at grain boundary might be a characteristic of Cr depletion by RIS because a width of Cr depletion at grain
166
T. Tsukada et al. / IASCC of type 316 SS irradiated in FBR
Fig. 10. Ductile fracture surface of the specimen aft& SSRT test carried out at 300°C in air. boundary on the neutron irradiated specimen was reported in the range of a few nanometer [12]. Dislocations are known to absorb radiation-produced interstitials preferentially during irradiation. This causes vacancies to flow to neutral sinks, such as cavities and grain boundaries. Vacancy flux to cavities causes cavity growth that leads to swelling of alloy and Ni-enrichment and Cr-depletion at cavity surfaces. The flow of vacancies to grain boundaries may also cause the same type of segregation during irradiation. Neutron spectrum of ORR was adjusted to achieve He/dpa ratio relevant to the fusion reactor first wall for Nibearing austenitic stainless steels [13]. The He/dpa
Potential (mV vs. SCE) Fig. 11. Polarization reactivation curves for FBR duct material obtained from EPR tests in standard and modified test solutions.
ratio of about 12.5 for ORR is larger than that of about 0.4 for JOY0 irradiation. Stoller et al. [14] indicated that vacancy flow to neutral sinks becomes maximum at a He/dpa ratio close to that expected in the fusion neutron environment. Sawai et al. [15] also reported that relatively large swelling of 0.3% occurred by ORR spectrally-tailored irradiation to about 8 dpa even at a low temperature of 400°C. Larger vacancy flux during ORR spectrally-tailored irradiation may cause larger swelling and more pronounced enrichment of Ni and depletion of Cr at grain boundaries. Swelling found in the fuel pin cladding irradiated in a similar condition with the duct material in this study was less than 0.1%. This suggests that segregation at grain boundaries in the duct material was smaller than that in the material irradiated in ORR. The duct material for JOY0 was cold worked before irradiation. Cold working introduces a large dislocation density. Although dislocations act preferential sinks for interstitial, strong sink strength of high density dislocations introduced by cold working reduces the vacancy concentration during irradiation and suppresses swelling. Therefore, cold working may also suppress the segregation and reduce IASCC susceptibility. This may also explain the dependence in IASCC susceptibility between the ORR and JOY0 irradiated materials. Under the LWR operation conditions, on the other hand, swelling does not occur because of lower fast neutron fluence and temperature. However, a susceptibility to IASCC of the material irradiated in LWR seems to be higher than that of the materials irradiated
T. Tsukada et al. / IASCC of type 316 SS irradiated in FBR
in ORR and JOY0 [3,4]. Further extension of these results to the LWR irradiation condition is difficult at present blut probably other effects, e.g. impurity segrerole for LWR gation, nnay also play an important materials
167
5. Conclusions Type 316 stainless steel from the core of tlle exl lerimental FBR “JOYO” was examined by the 5;SRT test in pure, oxygenated water and air and by the EPR test
Fig. 12. Specimen surfaces after EPR tests observed by SEM. (a) and (b) show nonuniform etching of grain faces and slight etching of grain boundaries resulted from standard EPR test. Cc) and (d) show uniform etching and narrow ditches formed at grain boundaries after modified EPR test.
168
i? Tsukada et al. / L4SCC
for the study of IASCC. The material was irradiated at 425°C to a neutron fluence of 8.3 x 102” n/m2 (> 0.1 MeV) which is equivalent to 40 dpa. The results are summarized as follows: (1) Intergranular cracking was observed on the fractured specimen tested by SSRT at 200 and 300°C in water. Area1 fractions of intergranular cracking portions in the fracture surface were increased with increasing test temperatures that were 0, 5 and 25% at 60, 200 and 3OO”C,respectively. (2) The specimen tested by SSRT at 300°C in air showed fully ductile fracture and, therefore, the intergranular cracking observed in water is considered as an evidence of IASCC. (3) Fracture of the specimens in high temperature water occurred by shearing between intergranular cracking portions caused by SCC. On the sheared surface, ductile-dimple pattern was observed. (4) After the EPR test, grain boundary etching was observed in addition to grain face etching. This may suggest Cr-depletions at grain boundaries and defect clusters. (5) Comparison with the SSRT test results from the spectrally-tailored irradiated type 316 material suggested a similar dependency between a susceptibility to IASCC and a swelling on He/dpa ratio. Differences of neutron spectrum and materials condition are considered to cause these dependencies.
Acknowledgements
The authors wish to acknowledge Dr. A. Hishinuma of JAERI for his helpful support and very grateful to staff members of the hot laboratories of JAERI and PNC. We are very much indebted to Prof. T. Shoji of Tohoku University for his valuable suggestions on the electrochemical test techniques and to Dr. G.E.C. Bell of Oak Ridge National Laboratory (ORNL) for his constructive advice. We also thank Drs. T. Kondo of JAERI, M. Katsuragawa of PNC and K. Tomabechi (formerly staff of JAERI) for their encouragement.
of type 316 SS irradiated in FBR References [l] J.L. Nelson and P.L. Andresen, Proc. 5th Int. Symp. on Environmental Degradation of Materials in Nuclear Power Systems - Water Reactors CANS, Monterey, California, 1991) p. 10. [2] H. Hanninen and I. Aho-Mantila, Proc. 3rd Int. Symp. on Environmental Degradation of Materials in Nuclear Power Systems - Water Reactors (AIME, Traverse City, Michigan, 1987) p. 77. [3] A.J. Jacobs, O.P. Wozadlo, K. Nakata, T. Yoshida and I. Masaoka, ibid. ref. [2], p. 673. [4] M. Kodama, S. Nishimura, J. Morisawa, S. Shima, S. Suzuki and M. Yamamoto, Proc. 5th Int. Symp. on Environmental Degradation of Materials in Nuclear Power Systems - Water Reactors CANS, Monterey, California, 19911 p. 948. [5] T. Tsukada, K. Shiba, G.E.C. Bell and H. Nakajima, Corrosion/92 (NACE, Nashville, TN, 1992) paper no. 104. [6] T. Tsukada, K. Shiba, M. Ohmi, M. Kizaki, H. Matsushima and H. Nakajima, Proc. 3rd Asian Symp. on Research Reactor (JAERI, Hitachi, Japan, 1991) p. 621. [7] K. Shiba, T. Tsukada, H. Nakajima, H. Matsushima, I. Takahashi, K. Sonobe and T. Komatsu, Japan Atomic Energy Research Institute Report, JAERI-M 91-024 (19911, in Japanese. [S] T. Inazumi, G.E.C. Bell, E.A. Kenik and K. Kiuchi, Corrosion 46 (10) (1990) 786. [9] ASTM standard G 108-92 (ASTM, 1992). [lo] Y. Iwabuchi, T. Fujimoto, Y. Watanabe and T. Shoji, Zairyo-to-Kankyo (Corrosion Engineering) 42 (1993) 2, in Japanese. [ll] T. Inazumi, G.E.C. Bell, P.J. Maziasz and T. Kondo, J. Nucl. Mater. 191-194 (1992) 1018. [12] E.A. Kenik, T. Inazumi and G.E.C. Bell, J. Nucl. Mater. 183 (1991) 145. [13] M.L. Grossbeck, E.E. Bloom, J.W. Woods, J.M. Vitek and K.R. Thorns, Proc. Conf. on Fast, Thermal and Fusion Reactor Experiments, vol. 1 CANS, Salt Lake City, Utah, 1982) pp. 1-199. [14] R.E. StoIIer, P.J. Maziasz and A.F. Rowcliffe, J. Nucl. Mater. 155-157 (1988) 1328. [15] T. Sawai, P.J. Maziasz, H. Kanazawa and A. Hishinuma, J. Nucl. Mater. 191-194 (1992) 712.