Effect of minor elements on irradiation assisted stress corrosion cracking of model austenitic stainless steels

Effect of minor elements on irradiation assisted stress corrosion cracking of model austenitic stainless steels

ELSEVIER Journal of Nuclear Materials 233-237 (1996) 1393-1396 jnurnalof nuclear materials Effect of minor elements on irradiation assisted stress ...

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ELSEVIER

Journal of Nuclear Materials 233-237 (1996) 1393-1396

jnurnalof nuclear materials

Effect of minor elements on irradiation assisted stress corrosion cracking of model austenitic stainless steels Y. Miwa a,., T. Tsukada a, S. Jitsukawa a, S. Kita a, S. Hamada a, Y. Matsui b, M. Shindo a a Department of Materials Science and Engineering. Japan Atomic Energy Research Institute. Tokai-mura, Naka-gun. lharaki-ken 319-I 1, Japan b Department of JMTR, Japan Atomic Energy Research Institute, Oarai-machi, Higasi-lbaraki-gun. lharaki-ken 311-11, Japan

Abstract A low impurity F e - 1 8 C r - 1 2 N i (HP) and its heats doped with Si and C (HP + Si and H P + C ) were irradiated to 6.7 × 10 24 n / m 2 ( E > 1 MeV) at 513 K. The slow strain rate tensile (SSRT) tests were carried out at a constant strain rate of 1.7 × 10 - 7 S I in high purity, 573 K water. Scanning electron microscopy on the fracture surface revealed that lIP and HP + Si failed mainly by the intergranular stress corrosion cracking (SCC), while the major failure mode in HP + C was the transgranular SCC. All alloys exhibited radiation hardening. HP + Si exhibiting the smallest hardening showed uniform elongation of 17%, while HP and HP + C did not. Transmission electron microscopy was also carried out. Frank loops and unidentified small clusters were formed in HP and HP + C, while only small clusters were observed in HP + Si.

1. Introduction Irradiation assisted stress corrosion cracking (IASCC) is one of the critical issues of austenitic stainless steels for the in-vessel structures of water-cooled fusion reactors [1], as well as those for the core components in light water reactors [2-7]. It is well known that for unirradiated thermally sensitized austenitic stainless steels, intergranular (IG) SCC in aqueous environments was induced by the Cr depletion at grain boundaries. Therefore, the Cr at grain boundaries occurred during irradiation are thought to also promote 1ASCC. On the other hand, there are some arguments about the role of Cr depletion for IASCC [8,9]. Recently, it was reported that radiation hardening also affected the susceptibility for 1ASCC. Effects of minor alloying elements on the radiation-induced microstructure, radiation hardening and fracture mode of IASCC are investigated.

* Corresponding author. Tel.: +81-29-2826082; fax: + 81-292825922; e-mail: [email protected]

2. Experimental procedure A low impurity F e - 1 8 C r - 1 2 N i (HP) and its heats doped with Si and C (HP + Si and liP + C) were used for the present study. The chemical compositions are shown in Table 1. These alloys were solution-annealed at 1373 K for 0.5 h. Neutron irradiation was performed at 513 +_ 10 K in helium environment for 3925 h in the Japan Research Reactor No. 3 at Japan Atomic Energy Research Institute to a fluence of 6,7 × l 0 24 n / m 2 ( E > 1 MeV) and 3 . 0 × 1025 n / m 2 for a fast and a thermal neutron, respectively. After irradiation, the SSRT tests were conducted in high-purity, 573 K water containing 32 ppm dissolved oxygen at 9.3 MPa with a flow rate of 5 l / h , Details of the test method were reported elsewhere [10]. SSRT test specimens were round bar type with a diameter of 4 mm and a gage length of 24 mm. The initial strain rate was 1.7 × l 0 - 7 S I. Electric conductivities of inlet water and outlet water were < 0,2 p~S/cm and 1.0-1.7 # S / c m , respectively. Fracture surfaces were examined in the scanning electron microscope (SEM) to evaluate the fracture mode, and the fraction of the SCC area was obtained.

0022-3115/96/$15.00 Copyright © 1996 Elsevier Science B.V. All rights reserved. PII S 0 0 2 2 - 3 1 1 5 ( 9 6 ) 0 0 2 5 5 - 3

Y. Miwa et al./Journal q/'Nu('lear MateriaLs 233-237 (1996) 1393 1396

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Table I Chemical compositions of the specimen materials (wt%)

HP HP ~ Si HP + C

C

Si

Mn

P

S

Cr

Ni

Ti

AI

N

Fe

0.003 0.003 0.098

0.01 0.69 0.03

1.36 1.36 1.39

0.001 0.001 0.00l

0.0014 0.0014 0.0020

I g. 17 18.01 18.30

12.27 12.24 12.50

0.01 < 0.00 < 0.00

0.16 0.10 0.11

0.0014 0.0014 0.0016

bal. hal. bal.

Microstructural observation was also carried out u s i n g the H F - 2 0 0 0 t r a n s m i s s i o n electron m i c r o s c o p e (TEM). T o reduce the g a m m a - r a y from the T E M disk s p e c i m e n , the c o m p o s i t e specinren technique [11] was applied. The activity o f the c o m p o s i t e s p e c i m e n was reduced to less than

1/40 of conventional specimens.

3. R e s u l t s

T h e s t r e s s - s t r a i n curves from the S S R T tests are s h o w n in Fig. I. O n the s t r e s s - s t r a i n curves for HP and HP + C. g(~)

--r--]

II]

7(KI 6(10

l

I

I~]

I

I

I

]

I

,

I

I

I

I

I

I~'~

Ruence: 6.7x102%/m 2 (E>IMeV) Irradiation temp.: 513 K -] SSRTtest temp.: 573 K ! Diss,,lved oxygen: 32 ppm

~ ] \ /

I

l

-

0

5

10

15

20

25

Strain (%) Fig. I. Nominal stress-storm curves from the SSRT tests in oxygenated high-purity water at 573 K.

stresses reached to m a x i n m m stress (¢r,1~,, ) without uniform elongation, while the t m i f o n n elongation was observed for HP + St. o'11~:,, o f HP + C is larger than lhat of HP. In HP + St. the yield stress is smaller than the ~r,..... of HP. but the ~5,m was about the saine as the o,,n:,~ of HP. In all alloys, intergranular cracking (IGC) and transgranular cracking ( T G C ) were observed. O n these IGC and T G C surfaces, s e c o n d a r y crocks were observed. B e c a u s e of this, these IGC and T G C surfaces are S C C surtimes. S E M i m a g e s are s h o w n in Fig. 2 ( a - c ) and the fractions of I G S C C area ( % I G S C C ) and T G S C C area ( g T G S C C ) are illustrated in Fig. 3. As seen in Fig. 2(a) and (b), liP and HP + Si failed mainly by I G S C C . and around this region small T G S C C and f o l l o w i n g ductile fracture were observed. B e c a u s e of b r a n c h i n g s and intersections o f cracks, I G S C C surface was not s m o o t h . The n u m b e r of s e c o n d a r y IGCs in HP + Si was h i g h e r than that in HP. % I G S C C in HP + Si was about the s a m e as that in HP (Fig. 3). However, it m u s t be noted that s p e c i m e n for liP + Si failed at the shoulder. O n the other hand, HP + C failed m a i n l y by T G S C C (Fig. 2(c)), though small I G S C C surface was formed before T G S C C occurred. T h e b n m c h i n g s of cracks on I G S C C surface were also found m HP + C. Furthermore, on the T G S C C surface o f HP + C, the fan shaped patterns and a n u n t b e r of short IGCs were also observed. T E M i m a g e s of radiation-induced microstructures are presented in Fig. 4 ( a - c ) . In all alloys, dislocation loops and small ctusters were observed, but irradiation-induced precipitates and depleted z o n e s of defects along grain boundaries were not observed. The size and n u m b e r den-

Fig. 2. SEM images of the SSRT lested specimens. 1GSCC are observed on HP (left) and HP + Si (cenler) and TGSCC tm HP - C (right).

Y. Miwa et al./Jaurnal o/'Nuclear Materials 233-237 (1996) 1393-1396

100 8O ~" 70

.~_ 40 ~ 3O m 20 10 0

"

HP

"

HP+Si

HP+C

Fig. 3. The fractions of IGSCC and TGSCC area on the fracture surface of specimens.

sity of defects were different from each alloy. In HP, Frank loops with an average diameter of 10 nm and small clusters less than 3 nm evolved. The number densities of Frank loops and visible small clusters were 9 × l0 21 m 3 and 5 × 10 22 m 3 respectively. In HP + Si, defects mainly consisted of small clusters. These small clusters were distributed homogeneously, while only a few small, localized Frank loops were recognized. The number density of Frank loops (NFLs) in HP + Si was not evaluated accurately. The total number density of observable small clusters and localized Frank loops in HP + Si was 4 × 10 22 m 3. In HP + C, the characteristic defect type was Frank loops with an average diameter of 7 rim. NFLs in HP + C is higher than that in HP, and was 6 × 10 22 m 3

4. D i s c u s s i o n

HP and HP + Si failed mainly by IGSCC and %IGSCC of both alloys were about 50%. HP + C exhibited only 4% of IGSCC and failed mainly by TGSCC with %TGSCC of about 68~. The %IGSCC at 6.7 x 10 24 n / m 2 ( E > 1

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MeV) for commercial purity 304 stainless steels have been described to be lower than 15% in several researches [3,12,13], indicating that the model alloys of the present experiment showed a higher susceptibility to IASCC comparing with commercial 304 stainless steels. Radiation hardening of HP + Si was significantly lower than that of HP. HP + Si also exhibited the largest postirradiation uniform elongation during SSRT test. This may be resulted from the lower hardening. HP + Si elongated to attain the maximum stress and failed by IGSCC during the test. The maximum stress level was comparable with that for HP, which failed also by IGSCC. The time and elongation to failure for HP + Si was, therefore, longer than those for HP. Failure by IGSCC after irradiation might be controlled by stress, and the smaller radiation hardening by Si addition resulted in a beneficial effect to retard SCC failure. HP + C with the highest flow stress level after irradiation failed mainly by TGSCC and exhibited only 4~7, of %IGSCC. This %IGSCC in HP + C was much smaller than %IGSCC in HP and HP + Si failed at the lower flow stress levels. Fukuya et al. [14] also reported smaller C~IGSCC for a heat with a higher flow stress level after irradiation, although they did not indicated that TGSCC had occurred. Tsukada et al. [15], however, reported that the failure mode changed from IGSCC to TGSCC occurred with a decrease in flow stress level by annealing after irradiation. The failure mode change did not seem to be controlled only by the flow stress level. In HP, Frank loops and small clusters were observed. The addition of C caused to increase in NFLs after irradiation, while the addition of Si almost completely suppressed the nucleation and the growth of the Frank loops. It is well known that Frank loops grow with absorbing radiation produced self interstitials preferentially, and this results in the increase of vacancy concentration to flow vacancies to neutral sinks, such as voids and grain boundaries. Growth of Frank loops enhanced by C addition, therefore, could

(a) HP

Fig. 4. Comparison of microstructures in irradiated HP (a), HP+ Si (b) and HP+ C (c). The beam direction is near < 1 1 0 > . 4' <0{)2>.

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Y. Miwa et al./Journal of Nuclear Materials 233-237 (1996) 1393-1396

result in the e n h a n c e m e n t o f radiation induced s e g r e g a t i o n (RIS) at grain boundaries to c h a n g e the m i c r o c h e m i s t r y , and could c a u s e to fail with IGSCC. H o w e v e r , H P + C tailed m a i n l y by T G S C C at a high stress level, but H P and HP + Si failed mainly by I G S C C at the relatively low stress levels. The relatively low irradiation temperature o f 513 K m i g h t limit the flow of radiation produced point defects across grains to grain boundaries, and RIS to grain boundaries m i g h t be caused by flow of the point defects in the region vicinity to grain boundaries. The m i c r o c h e m i c a l analysis m a y provide a key to u n d e r s t a n d the controlling factor to suppress I G S C C for HP + C.

5. Conclusion A low impurity F e - 1 8 C r - 1 2 N i and its heats d o p e d with Si and C were irradiated to 6.7 X 10 2`a n / m : ( E > 1 MeV). T h e S S R T tests were carried out after irradiation, and the fracture surfaces were e x a m i n e d in SEM. Microstructural observation has been carried out with T E M . Following c o n c l u s i o n s were drawn: 1. HP + Si s h o w e d a m u c h lower radiation h a r d e n i n g than HP, t h o u g h the fraction o f I G S C C area was as s a m e as that in HP. 2. HP + C s h o w e d a h i g h e r radiation h a r d e n i n g than HP, and failed m a i n l y by T G S C C . 3. T h e addition o f Si decreased the n u m b e r density of Frank loops, while the addition of C increased that. 4. No o b v i o u s relation was detected b e t w e e n the susceptibility for I G S C C and radiation p r o d u c e d microstructure.

Acknowledgements T h e authors would like to thank H. N a k a j i m a , for helpful d i s c u s s i o n and support, and the staff o f the Japan A t o m i c E n e r g y R e s e a r c h Institute hot laboratories, for p e r f o r m i n g S S R T tests and other e x a m i n a t i o n s .

References [1] IAEA, Research and Development Needs ff)r ITER Engineering Design, I T E R / D S / N o . 20 (IAEA, 1991).

[2] A.J. Jacobs, G.P. Wozadlo, K. Nakata, T. Yoshida and 1. Masaoka, Proc. Third lnt. Syrup. Environ. Degradation of Mater. in Nucl. Power Syst.-Water Reactors (The Metallurgical Society, Traverse City, MI, 1988) p. 673. [3] M. Kodama, S. Nishimura, J. Morisawa, S. Suzuki, S. Shima and M. Yamamoto, Proc. Fifth Int. Syrup. Environ. Degrada tion of Mater. in Nuel. Power Syst. Water Reactors (American Nuclear Society, Montrey, CA, 1992) p. 948. [4] R. Katsura, J. Morisawa, M. Kodama, S. Nishimura, S. Suzuki, S. Shima and M. Yamamoto, Proc. Sixth Int. Syrup. Environ. Degradation of Mater. in Nucl. Power Syst. Water Reactors (The Minerals, Metals and Materials Society, San Diego, CA, 1993) p. 625. [5] A.J. Jacobs, G.P. Wozadlo and S.A. Wilson, Corrosion 4~2) (1993) p. 145. [6] S. Kasahara, K. Nakata, A.J. Jacobs, G.P. Wozadlo, K. Fukuya, S. Shima and S. Suzuki, Proc. Fifth Int. Symp. Environ. Degradation of Muter. in Nucl. Power Syst.-Watcr Reactors (American Nuclear Society, Montrey, CA, 1992) p. 615. [7] F. Garzarolli, D. Alter, P. Dewes and J.L. Nelson, Proc. Third Int. Syrup. Environ. Degradation of Mater. in Nucl. Power Syst.-Water Reactors (The Metallurgical Society, Traverse City, MI, 1988) p. 657. [8] E.D. Eason and E.E, Nelson, Proc. Seventh Int. Syrup. Environ. Degradation of Mater. in Nucl. Power Syst.-Water Reactors (Anrerican Nuclear Society, Breckenridge, CO, 1995), to be published. [9] H.M. Chung, W.E. Ruther, J.E. Sanecki, A. Hins and T.F. Kassner, Proc. Seventh Int. Syrup. Environ. Degradation of Mater. in Nucl. Power Syst.-Water Reactors (American Nuclear Society, Brcckenridgc, CO, 1995), to be published. [10] T. Tsukada, K. Shiba, G.E.C. Bell and H. Nakajima, Corro sion/92, Paper No. 104 (National Association of Corrosion Engineers, Houston, TX, 1992). [11] S. Hamada and K. Hojou, J. Nucl. Mater. 200 (1993) 149. [12] A.J. Jaeobs, D. A. Hale and M. Sicglcr, GE Nuclear Energy, San Jose, CA, January 1986, SCC data as published in Ref. [2]. [13] W.L. Clarke and A.J. Jacobs, Corrosion 1983 (NACE, 1983) p. 451. [14] K. Fukuya, K. Nakata, S. Kasahara, A.J. Jacobs, G.P. Wozadlo, S. Suzuki and M. Kitamura, Proc. Sixth Int. Synrp. Environ. Degradation of Mater. in Nucl. Power Syst.-Water Reactors (The Minerals, Metals and Materials Society, San Diego, CA, 1993) p. 565. [15] T. Tsukada and H. Nakajima, J. Nucl. Mater. 212 215 (1994) 1519.