SNUREG/CR-6687

ANL-00/21

Irradiation-Assisted Stress Corrosion Cracking of Model Austenitic Stainless Steel Alloys

Argonne National Laboratory

U.S. Nuclear Regulatory Commission Office of Nuclear Regulatory Research Washington, DC 20555-0001

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NUREG/CR-6687 ANL-00/21

Irradiation-Assisted Stress Corrosion Cracking of Model Austenitic Stainless Steel Alloys

Manuscript Completed: May 2000 Date Published: October 2000

Prepared by H. M. Chung, W. E. Ruther, R. V. Strain, W. J. Shack

Argonne National Laboratory 9700 South Cass Avenue Argonne, IL 60439

M. McNeil, NRC Project Manager

Prepared for Division of Engineering Technology Office of Nuclear Regulatory Research U.S. Nuclear Regulatory Commission Washington, DC 20555-0001 NRC Job Code W6610

Irradiation-Assisted Stress Corrosion Cracking of Model Austenitic Stainless Steel Alloys

by H. M. Chung, W. E. Ruther, R. V. Strain, and W. J. Shack

Abstract

This report summarizes work performed at Argonne National Laboratory on irradiation

assisted stress corrosion cracking (IASCC) of model austenitic stainless steels (SSs) that were

irradiated in the Halden boiling heavy water reactor in simulation of the neutronic and

environmental conditions of boiling water reactor (BWR) core internal components. Slow

strain-rate tensile tests in simulated BWR water were conducted on model austenitic stainless

steel alloys that were irradiated at 2890C in helium to =0.3 x 1021 n-cm-2 and =0.9 x 1021

n-cm- 2 (E > 1 MeV). Fractographic analysis by scanning electron microscopy was conducted

to determine susceptibility to irradiation-assisted stress corrosion cracking as manifested by

the degree of intergranular and transgranular stress corrosion cracking (IGSCC and TGSCC)

fracture surface morphology. As fluence was increased from -0.3 x 1021 n-cm-2 to =0.9 x 1021

n-cm-2 , IGSCC fracture surfaces emerged in many alloys, usually in the middle of and

surrounded by TGSCC fracture surfaces. Alloy-to-alloy variations in susceptibility to TGSCC

and IGSCC were significant at -0.9 x 1021 n-cm-2 . Susceptibility to TGSCC and IGSCC was

influenced by more than one alloying and impurity element in a complex manner. Results

from this study indicate that for commercial heats of Types 304 and 304L SS, a high

concentration of S is detrimental and that a sufficiently low concentration of S is a necessary

condition for good resistance to IASCC. A laboratory alloy containing a high concentration of

Cr exhibited excellent resistance to TGSCC and IGSCC, despite a high S content, indicating

that Cr atoms in high concentration play a major role in suppressing susceptibility to IASCC

under BWR conditions. Yield strength of the model alloys, measured in BWR-like water at

2890C, was nearly constant at =200 MPa in the unirradiated state and was more or less

independent of Si concentration. However, as the alloys were irradiated, the degree of increase

in yield strength was significantly lower for alloys that contain >0.9 wt.% Si than for alloys

that contain <0.8 wt.% Si, indicating that the nature of irradiation-induced hardening centers

and the degree of irradiation hardening is significantly influenced by the Si content of the

alloy. Similar influence was not observed for C and N. Results also indicate that a Si content

between -0.8 and -1.5 wt.% is beneficial in delaying the onset of or suppressing the

susceptibility to IASCC. Although susceptibility to TGSCC and IGSCC was insignificant and

the fracture surface morphology was mostly ductile, some alloys exhibited very low uniform

elongation and poor work-hardening capability in water after irradiation to -0.9 x 1021 n-cm-2 .

Such alloys contained unusually high concentrations of 0 or unusually low concentrations of

Si or both. To better understand such behavior, usually indicative of severe susceptibility to

dislocation channeling or other forms of localized deformation, a systematic microstructural

investigation by transmission and scanning electron microscopies is desirable.

NUREG/CR-6687

Contents

Executive Summary .......................................................................................................... ix

Acknowledgments ............................................................................................................. xi

1 Introduction ........................................................................................................ 1

2 Test Matrix and Specimen Fabrication .................................................................... 3

2.1 Construction of Test Matrix ........................................................................... 3

2.2 Fabrication of Test Specimens ........................................................................ 6

3 Specimen Irradiation ............................................................................................... 7

4 SSRT Test and SEM Fractography on Medium-Fluence Specimens .......................... 9

4.1 Results of SSRT Tests and SEM Fractography ................................................ 9

4.2 Effect of Sulfur ........................................................................................... 20

4.3 Effect of Chromium ...................................................................................... 23

4.4 Effect of Silicon ............................................................................................... 23

4.5 W ork-Hardening Capability .......................................................................... 26

5 Conclusions ............................................................................................................. 27

References ....................................................................................................................... 29

-v- NUREG/CR-6687

Figures

1. Geometry of slow-strain-rate tensile test specimens ......................................... 6

2. Effects of fluence on yield strength measured in 289°C water containing =8 ppm DO ............................................. 16

3. Effects of fluence on maximum strength measured in 289°C water containing =8 ppm DO ................................................................................... 16

4. Effects of fluence on uniform elongation measured in 289°C water containing =8 ppm DO ................................................................................... 17

5. Effects of fluence on total elongation measured in 289°C water containing -8 ppm DO ................................................................................... 17

6. Effects of fluence on percent TGSCC measured in 2890C water containing -8 ppm DO ................................................................................... 18

7. Effects of fluence on percent IGSCC measured in 2890C water containing =8 ppm DO ................................................................................... 18

8. Effects of fluence on percent TGSCC + IGSCC measured in 2890C water containing -8 ppm DO .......................................................................... 19

9. Effects of fluence and test environment on load elongation behavior of Type 304 SS commercial heat C19 ................................................................. 19

10. Effect of S on susceptibility to TGSCC of two groups of alloys in unirradiated state or after irradiation to -0.3 x 1021 n.cm- 2 (E > 1 M ev) ............................................................................................................... 20

11. Effect of S on susceptibility to TGSCC and IGSCC of commercial alloys measured after irradiation to =0.9 x 1021 n-cm-2 (E > 1 MeV) ................ 21

12. Effect of S on uniform and total elongation of commercial alloys measured after irradiation to -0.9 x 1021 n-cm- 2 (E > 1 MeV) ........................ 22

13. Effect of Cr on susceptibility to TGSCC and IGSCC of laboratory alloys measured after irradiation to -0.9 x 1021 ncm-2 (E > I MeV) ................ 24

14. Effect of Si concentration on yield strength of Types 304 and 304L alloys measured in 2890C water before and after irradiation .......................... 25

15. Effect of Si on susceptibility to IGSCC of laboratory alloys of Types 304 and 304L SSs measured after irradiation to =0.9 x 1021 n-cm- 2 (E > I M eV) ........................................................................................................ 25

NUREG/CR-6687 - vi -

Tables

1. Compositions assigned to medium, high, and low levels of alloying and

impurity elements used to construct Taguchifs L18(21 x 37) array ..................... 3

2. Basic test matrix optimized according to Taguchi's orthogonal array

L 18 (2 1 x 37 ) .......................................................................................................... 4

3. Composition of six supplementary commercial- and high-purity heats

of Types 304, 316, and 348 stainless steels ...................................................... 4

4. Elemental composition of 27 commercial and laboratory model

austenitic stainless steel alloys irradiated in Halden Reactor ............................. 5

5. Summary of specimen number per alloy, irradiation fluence, and

postirradiation test type ................................................................................... 7

6. Summary of discharge fluence of model austenitic stainless steel

alloys irradiated in the Halden Boiling Heavy Water Reactor ............................. 8

7. Results of SSRT test and SEM fractography for nonirradiated control

specimens of model austenitic stainless steel alloys ......................................... 10

8. Compositional characteristics of nonirradiated control specimens of

model austenitic stainless steel alloys correlated with results of SSRT

tests and SEM fractography ............................................................................ 11

9. Results of SSRT tests and SEM fractography for model austenitic

stainless steels irradiated in helium at 2890C to fluence of =0.3 x 1021

n cm -2 . . . . . . . . .................................................. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 12

10. Compositional characteristics of model austenitic stainless steels irradiated to fluence of =0.3 x 1021 n cm-2 correlated with results of

SSRT tests and SEM fractography .................................................................. 13

11. Results of SSRT test and SEM fractography for model austenitic

stainless steels irradiated in He at 289°C to fluence of =0.9 x 1021

n cm -2 . . . . . . . . .................................................. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 14

12. Compositional characteristics of model austenitic stainless steels

irradiated to fluence of =0.9 x 1021 n cm-2 correlated with results of

SSRT test and SEM fractography .................................................................... 15

NUREG/CR-6687- vii -

Executive Summary

This report summarizes experimental work performed at Argonne National Laboratory on

irradiation-assisted stress corrosion cracking of model austenitic stainless steels that were

irradiated in the Halden Boiling Heavy Water Reactor in simulation of irradiation-induced

degradation of boiling water reactor (BWR) core internal components. Slow-strain-rate tensile

tests in simulated BWR water were conducted on twenty-three model austenitic stainless steel

alloys that were irradiated at 288°C in helium in the Halden reactor to a fluence of -0.9 x 1021

n.cm- 2 (E > 1 MeV). A similar investigation of the behavior of sixteen alloys irradiated to a

fluence of -0.3 x 1021 n-cm-2 (E > I MeV) has been reported previously. Fractographic

analysis by scanning electron microscopy was conducted to determine the susceptibility to

irradiation-assisted stress corrosion cracking as manifested by the degree of intergranular and

transgranular fracture surface morphology.

At -0.3 x 1021 n-cm-2 (E > 1 MeV), many alloys were susceptible to transgranular stress

corrosion cracking, but only one alloy of Type 316L stainless steel that contains a very low

concentration of Si (0.024 wt.% Si) was susceptible to intergranular stress corrosion cracking.

Alloy-to-alloy variations in susceptibility to transgranular stress corrosion cracking were

significant at =0.3 x 1021 n-cm- 2 . As fluence was increased from =0.3 x 1021 n-cm-2 to -0.9 x 1021 n-cm- 2 , intergranular fracture surfaces emerged in many alloys, usually in the middle of

and surrounded by transgranular fracture surfaces. This indicates that for many alloys high

susceptibility to transgranular stress corrosion cracking is a precursor to susceptibility to

intergranular stress corrosion cracking at a higher fluence. Alloy-to-alloy variations in

susceptibility to transgranular and intergranular stress corrosion cracking were significant at

=0.9 x 1021 n-c in-2 . Susceptibility to transgranular and intergranular stress corrosion

cracking was influenced by more than one alloying and impurity element in a complex

manner.

In unirradiated state and at the relatively low fluence level of =0.3 x 1021 n-cm- 2 ,

commercial and laboratory heats of Types 304, 304L, and 348 stainless steel that contain

relatively high concentrations of S (>0.009 wt.% S, 15-18.50 wt.% Cr) exhibited significant

susceptibility to transgranular stress corrosion cracking, whereas alloys that contain relatively

low concentrations of S (<0.005 wt.% S) exhibited good resistance to transgranular stress

corrosion cracking. At =0.9 x 1021 n-cm-2 , commercial heats of Types 304 and 316 stainless

steel that contain low concentrations of S exhibited high resistance to transgranular and

intergranular stress corrosion cracking and high levels of uniform and total elongation,

whereas a commercial heat of Type 304 stainless steel that contained a relatively high

concentration of S exhibited poor resistance to transgranular and intergranular stress

corrosion cracking and low levels of uniform and total elongation. These observations indicate

that for commercial heats of Types 304 and 304L stainless steel, a high concentration of S

(>0.009 wt.% S) is significantly detrimental and that a sufficiently low concentration of S is a

necessary condition for good resistance to IASCC. At -0.9 x 1021 n-cm-2 , a laboratory alloy

that contains a high concentration of Cr (-21 wt.%) exhibited excellent resistance to

transgranular and intergranular stress corrosion cracking, in spite of a high S content (-0.028

wt.% S), whereas heats with <17.5 wt.% Cr exhibited significant susceptibility to transgranular

plus intergranular stress corrosion cracking. This finding indicates that Cr atoms in high

concentration play a major role in suppressing susceptibility to irradiation-assisted stress

corrosion cracking under BWR conditions.

NUREG/CR-6687- ix -

Yield strength of unirradiated model alloys, measured in BWR-like water at 2890C, was nearly constant at =200 MPa and was more or less independent of Si concentration. However, as the alloys were irradiated to -0.3 x 1021 n-cm- 2 and -0.9 x 1021 n-cm- 2 , the degree of increase in yield strength was significantly lower for alloys that contain >0.9 wt.% Si than for alloys that contain <0.8 wt.% Si. This observation indicates that the nature of irradiationinduced hardening centers and the degree of irradiation hardening is significantly influenced by the Si content of the alloy. A similar influence was not observed for C and N. Among laboratory heats of Types 304 and 304L stainless steel, alloys that contain <0.67 wt.% Si exhibited significant susceptibility to intergranular stress corrosion cracking, whereas heats with 0.8-1.5 wt.% Si exhibited negligible susceptibility to intergranular stress corrosion cracking. However, an alloy with --1.9 wt.% Si exhibited some degree of susceptibility to intergranular stress corrosion cracking. These observations indicate that a Si content between -0.8 and =--1.5 wt.% is beneficial in delaying the onset of, or suppressing, susceptibility to irradiation-assisted stress corrosion cracking. Consistent with the effects of S, Cr, and Si, an alloy that contains unusually low concentrations of Cr and Si and unusually high concentrations of S and 0 exhibited the worst susceptibility to IASCC at -0.9 x 1021 n-cm- 2 .

Although susceptibility to transgranular and intergranular stress corrosion cracking was insignificant and the fracture surface morphology was mostly ductile, some alloys exhibited very low uniform elongation and poor work-hardening capability in water after irradiation to =0.9 x 1021 n-cm-2 . Such alloys contained unusually high concentrations of 0 or unusually low concentrations of Si or both. To better understand such behavior, usually indicative of severe susceptibility to dislocation channeling or other forms of localized deformation, a systematic microstructural investigation by transmission and scanning electron microscopies will be needed. A few alloys that contain unusually high concentrations of 0 exhibited either high susceptibility to intergranular stress corrosion cracking or poor work-hardening capability. Therefore, it appears that a high concentration of 0 is also detrimental.

NUREG/CR-6687

Acknowledgments

The authors sincerely thank T. M. Karlsen, OECD Halden Reactor Project, Halden,

Norway, for performing specimen irradiation in the Halden reactor; D. R. Perkins and R. W.

Clark for conducting the slow-strain-rate tensile tests on the irradiated specimens. The

authors are also grateful to M. B. McNeil for many helpful discussions.

NUREG/CR-6687-xid-

I Introduction

In recent years, failures of reactor core internal components have increased after

accumulating a fluence >0.5 x 1021 n-cm- 2 (E >1 MeV), or -0.7 dpa, in boiling water reactors (BWRs), and at approximately one order of magnitude higher fluences in some pressurized water reactors (PWRs). The general pattern of the observed failures indicates that as nuclear

plants age and neutron fluence increases, various nonsensitized austenitic stainless steels

(SSs) become susceptible to intergranular failure. Some components are known to have

cracked under minimal applied stress. Although most failed components can be replaced,

some safety-significant structural components (e.g., the BWR top guide, shroud, and core

plate) would be very costly or impractical to replace. Therefore, the structural integrity of

these components at high fluence has been a subject of concern, and extensive research has been conducted to provide an understanding of this type of degradation, which is commonly known as irradiation-assisted stress corrosion cracking (IASCC).

Irradiation profoundly affects both local coolant-water chemistry and component

microstructure. The primary effects of irradiation on materials include alteration of local

microchemistry, microstructure, and mechanical properties of the core internal components, which are usually fabricated from American Society of Testing and Materials (ASTM) Types

304, 316, or 348 SS. Irradiation produces defects and defect clusters in grain matrices, alters

the dislocation and dislocation loop structures, and produces defect-impurity and defectcluster-impurity complexes, leading to radiation-induced hardening and, in many cases, flow

localization via dislocation channeling. Irradiation also leads to changes in the stability of

second-phase precipitates and the local alloy chemistry near grain boundaries, precipitates, and defect clusters. Grain-boundary microchemistry that significantly differs from bulk

composition can be produced in association with not only radiation-induced segregation but

also thermally driven equilibrium and nonequilibrium segregation of alloying and impurity elements.

For many years, irradiation-induced grain-boundary depletion of Cr has been considered the primary metallurgical process that causes IASCC. One of the most important factors that has been considered by many investigators to support the Cr-depletion mechanism is the similarity in dependence on water chemistry (i.e., oxidizing potential) of intergranular stress corrosion cracking (IGSCC) of both nonirradiated thermally sensitized material and IASCC of

BWR-irradiated solution-annealed material. 1-3 However, contrary to expectations based on

the Cr-depletion mechanism, cracking incidents of control rod cladding and baffle plate bolts

have been reported at numerous PWRs (i.e., under nonoxidizing potential). Also, the susceptibility of PWR-irradiated components to IASCC has been shown clearly by expandingmandrel4 and slow-strain-rate tensile (SSRT)5 tests in PWR water4 or PWR-simulated water,5

although PWR water chemistry falls well within the range of protective electrochemical potential (ECP).1-3

Other investigators have implicated radiation-induced segregation of ASTM-specified impurities, such as Si, P, and S, as the primary process that causes IASCC.4 ,6,7 The superior resistance of one heat of Type 348 SS that is especially low in C, Si, P, and S seemed to

provide support for this process, 4 and the same rationale was extended to Type 304 SS. However, in direct contradiction, several investigators recently reported that resistance of

high-purity (HP) heats (low in C, Si, S, and P) of Type 304 SS is no better than that of

NUREG/CR-6687-1

commercial-purity (CP) Type 304 SSs.8 -14 Therefore, the role of grain-boundary segregation of Si, P, and S appears to be not well established.

Although C significantly increases the yield strength of irradiated stainless steels, higher C content seems to be either benign or even conducive to lower susceptibility to intergranular cracking of irradiated materials. 14 Deleterious effects of 0 in steels have been reported by Chung et al. 14 and Cookson et al.1 5 Indications of the deleterious effect of grain-boundary segregation of N have been reported for BWR neutron absorber tubes fabricated from highand commercial-purity heats of Type 304 SS.14 Similar reports suggest that a higher concentration of N is deleterious, at least under BWR conditions.6,12,16, 17 Indications of the deleterious role of N have also been reported for Types 304L and 316L SSs that contain <0.024 wt.% C and have been irradiated in BWRs or test reactors at 240-300C.11 Kasahara et al.16 also reported that higher N in Type 316L increased the susceptibility to IASCC, indicating that Type 316LN is a susceptible material. This observation is consistent with the behavior of 316NG SSs reported by Jacobs et al. 6 and Jenssen and Ljungberg. 1 7 In contrast to these observations, 316NG SSs irradiated at -50°C has been reported to be resistant to intergranular failure at -288°C in water that contained =32 ppm dissolved oxygen (DO).11 Therefore, the role of N appears to be neither clear nor convincing.

In general, IASCC is characterized by strong heat-to-heat variation in susceptibility, in addition to strong dependence on irradiation condition, material type, and grade, even among materials of virtually identical chemical composition. This situation indicates that the traditional interpretation based on the role of grain-boundary Cr depletion cannot completely explain the mechanism of IASCC. Thus, although most investigators believe that significant grain-boundary Cr depletion plays an important role, other important processes that could be associated with other minor impurity elements may have been overlooked. 14 Therefore, we have initiated a new irradiation testing program to systematically investigate the effects of alloying and impurity elements (Cr, Ni, Si, P, S, Mn, C, and N) on the susceptibility of austenitic SSs to IASCC at several fluence levels. Construction of the test matrix was based on the method of Taguchi, 18 and susceptibility to IASCC was determined by SSRT testing of the specimens irradiated in BWR-simulated water and post-testing fractographic examination by scanning electron microscopy (SEM). This report summarizes the results obtained from 16 model austenitic stainless steel alloys that were irradiated at 288°C in He in the Halden reactor to a fluence of =0.3 x 1021 n-cm-2 (E > 1 MeV) and 23 alloys irradiated to a fluence of -0.9 x 1021 n-cm- 2 (E > 1 MeV).

NUREG/CR-6687 -2-

2 Test Matrix and Specimen Fabrication

2.1 Construction of Test Matrix

The irradiation test matrix was constructed according to the method of Taguchi, which is

described in Ref. 18. The base matrix followed Taguchi's standard orthogonal array L18 (21 x

37) which is an optimized matrix designed to determine systematically the effects of seven

variables (i.e., bulk material concentrations of Cr, Si, P, S, Mn, C, and N) at three

concentration levels, and one variable (Ni concentration) at two levels. A possible synergistic

interaction was assumed only between Ni and Si. Compositions assigned to the three levels of

the seven alloying and impurity elements and two levels of Ni are given in Table 1. The basic

test matrix, optimized on the basis of Taguchi's orthogonal array L18 (21 x 37), is given in

Table 2.

Table 1. Compositions assigned to medium, high, and

low levels of alloying and impurity elements used to construct Taguchi's Li 8 (21 x 37)

array

Composition (in wt.%)

Factor Element Assigned to Level

No. Assigned High Medium Low

1 Ni 10.000 8.000

2 Si 1.400 0.700 0.150

3 P 0.090 0.040 0.020

4 S 0.020 0.006 0.002

5 Mn 2.050 1.200 0.500

6 C 0.090 0.040 0.010

7 N 0.090 0.040 0.010

8 Cr 18.500 17.500 16.000

In addition to the 18 statistically optimized heats given in Table 2, six supplementary

heats of commercial- and high-purity (CP and HP) grade Types 304, 316, and 348 SS were

included in the test matrix. Compositions of the six supplementary heats are given in Table 3.

Of the 24 heats given in Tables 1 and 2, eight (i.e., Heats 1, 3, 9, 10, 12, 16, 19, and 21) were

replaced by commercially fabricated and purchased heats. Compositions of major impurities

(i.e., Si, P, C, and N) of each of the eight commercial heats matched closely those of each

corresponding heat given in Tables 1 and 2. The prefix "C" was added to the identification

number of these eight commercial heats (see Table 4).

The remaining 16 heats in Tables 1 and 2 were fabricated in the laboratory. All of the

laboratory-fabricated heats were designated with identification numbers that began with "LU.

To this matrix of 24 heats, three laboratory heats were added to test the effects of the

fabrication procedure. Compositions of these three laboratory heats (i.e., Heats L25C3,

NUREG/CR-6687-3-

L26C 19, and L27C21 in Table 4) closely match those of the corresponding commercial heats

(Heats C3, C19, and C21), respectively. Elemental compositions of the complete test matrix,

comprising 27 model austenitic SSs, is given in Table 4.

Table 2. Basic test matrix optimized according to Taguchi's orthogonal array L18 (21 x 3 7)

Heat Composition (in wt.%o)

Ni Si P S Mn

10.0 0.70 0.04 0.006 1.20 10.0 0.70 0.09 0.020 2.05 10.0 0.70 0.02 0.002 0.50

10.0 1.40 0.04 0.006 2.05 10.0 1.40 0.09 0.020 0.50 10.0 1.40 0.02 0.002 1.20

10.0 0.15 0.04 0.020 1.20 10.0 0.15 0.09 0.006 2.05 10.0 0.15 0.02 0.002 0.50

8.0 0.70 0.04 0.002 0.50 8.0 0.70 0.09 0.006 1.20 8.0 0.70 0.02 0.020 2.05

8.0 1.40 0.04 0.020 0.50 8.0 1.40 0.09 0.002 2.05 8.0 1.40 0.02 0.006 1.20

8.0 0.15 0.04 0.002 2.05 8.0 0.15 0.09 0.006 0.50 8.0 0.15 0.02 0.020 1.20

C 0.04 0.09 0.01

0.09 0.01 0.04

0.01 0.04 0.09

0.09 0.01 0.04

0.04 0.09 0.01

0.01 0.04 0.09

N Cr 0.04 18.5 0.09 17.5 0.01 16.0

0.01 16.0 0.04 18.5 0.09 17.5

0.09 16.0 0.01 18.5 0.04 17.5

0.09 18.5 0.01 17.5 0.04 16.0

0.01 17.5 0.04 16.0 0.09 18.5

0.04 17.5 0.09 16.0 0.01 18.5

Table 3. Composition (in wt.%) of six supplementary commercial- and high-purity

(CP and HP) heats of Types 304, 316, and 348 stainless steels

Heat SS Type C N Si P S Mn Cr Ni Nb Mo

19 CP 304 0.04 0.04 0.70 0.04 0.06 1.20 18.5 10.0

20 HP 304 0.01 0.01 0.15 0.02 0.02 0.50 18.5 10.0 -

21 CP 316 0.04 0.04 0.70 0.04 0.06 1.20 16.5 13.5 - 2.20

22 HP 316 0.01 0.01 0.15 0.02 0.02 0.50 16.5 13.5 - 2.20

23 CP 348 0.04 0.04 0.70 0.04 0.06 1.20 17.5 11.5 2.20

24 HP 348 0.01 0.01 0.15 0.02 0.02 0.50 17.5 -11.5 2.20 -

NUREG/CR-6687

No. 1

2 3

4 5 6

7 8 9

10 11 12

13 14 15

16 17 18

-4-

Composition (in wt.%)

Table 4. Elemental composition of 27 commercial and laboratory model austenitic

stainless steel alloys irradiated in Halden Reactor

ANL Source Composition (wt.%)

fDa Heat ID Ni Si P S Mn C N Cr 0 B Mo or Nb

, ~ ~ ~ n3 A o1fir lf nnV) r%~f' 1AA' % nA~ nAA n 1I -I I n1 -%,I

L2

C3

IA

L5

L6

L7

L8

C9

jI'Pt.- i V, Io

BPC-4-111

PNL-C-1

BPC-4-88

BPC-4-104

BPC-4-127

BPC-4-112

BPC-4-91

PNL-C-6

C10 DAN-23381

L11 BPC-4-93

C12 DAN-23805

L13 BPC-4-96

L14 BPC-4-129

L15 BPC-4-126

C16 PNL-SS-14

L17 BPC-4-128

L18 BPC-4-98

C19

L20 C21b

L22C

L23d

L24e

L25C3 L26C19

L27C21

DAN-74827

BPC-4-101

DAN- 12455

BPC-4-100

BPC-4-114

BPC-4-105

BPC-4-133

BPC-4-131

BPC-4-132

O. .LZ U.OU U.U

10.50 0.82 0.080

8.91 0.46 0.019

10.20 0.94 0.031

9.66 0.90 0.113

10.00 1.90 0.020

10.60 0.18 0.040

10.20 0.15 0.093

8.75 0.39 0.013

8.13 0.55 0.033

8.15 0.47 0.097

8.23 0.47 0.018

8.18 1.18 0.027

7.93 1.49 0.080

8.00 1.82 0.010

12.90 0.38 0.014

8.00 0.66 0.090

8.13 0.14 0.016

8.08 0.45 0.031

8.91 0.17 0.010

10.24 0.51 0.034

13.30 0.24 0.015

12.04 0.68 0.030

12.30 0.03 0.007

8.93 0.92 0.020

8.09 0.79 0.004

10.30 0.96 0.040

0.034

0.004

0.010

0.028

0.005

0.038

0.010

0.013

0.002

0.009

0.002

0.022

0.002

0.013

0.002

0.009

0.033

0.003

0.004

0.001

0.004

0.047

0.005

0.008

0.002

0.002

1.58 0.074

1.81 0.016

1.75 0.110

0.47 0.006

1.13 0.096

1.02 0.007

1.85 0.041

1.72 0.062

1.00 0.060

1.02 0.014

1.00 0.060

0.36 0.026

1.76 0.107

1.07 0.020

1.66 0.020

0.48 0.061

1.13 0.080

0.99 0.060

0.41 0.002

1.19 0.060

0.40 0.003

0.96 0.043

0.48 0.031

1.54 0.019

0.91 0.070

0.97 0.057

0.102

0.083

0.002

0.033

0.087

0.111

0.001

0.065

0.086

0.004

0.070

0.001

0.028

0.085

0.011

0.078

0.001

0.070

0.002 0.020

0.001

0.092

0.002

0.095

0.089

0.019

17.02

18.55

15.80

21.00

17.10

15.40

18.30

18.48

18.19

17.40

18.43

17.40

15.00

17.80

16.92

15.30

18.00

18.21

18.10 16.28

16.10

17.30

16.90

17.20

17.20

15.30

0.0066 <0.001

<0.001

- <0.001 - <0.001

0.0058 <0.001

0.0274 <0.001

- <0.001

- <0.001

<0.001 <0.001

<0.001

- <0.001 0.0045 <0.001

0.0110 <0.001

- <0.001

0.0090 <0.001

- <0.001

- <0.001

- <0.001

- <0.001

- <0.001

0.0093 <0.001

0.0129 <0.001

0.0085 0.0080

0.0058

0.010 <0.001

0.030

aFirst letters "C" and "L" denote commercial and laboratory heats, respectively. bCommercial-purity Type 316 SS. cHigh-purity Type 316 SS. dCommercial-purity Type 348 SS. eIIgh-purity Type 348 SS.

NUREG/CR-6687

Mo 2.08

Mo 2.04

Nb 1.06

Nb 1.72

Mo 2.01

-5-

2.2 Fabrication of Test Specimens

Slow-strain-rate tensile specimens were machined from solution-annealed and waterquenched plates or sheets that were fabricated from the 27 model austenitic SS alloys. The geometry of the SSRT specimens, 0.76 mm thick, 57.2 mm long, and 12.7 mm wide (gauge section 19.1 mm long and 3.1 mm wide), is shown in Fig. 1. Gauge lengths and planes of the specimens were parallel to the rolling direction and plane of the sheets, respectively. Subsize compact-tension (1/4TCT) specimens were also irradiated in tandem in the same capsules with the SSRT specimens to determine J-R fracture toughness properties and crack growth rates (CGRs) after irradiation. Together, 96 SSRT and 24 CT specimens were prepared and irradiated in this study. After these specimens were mechanically machined in the shop, no additional heat-treatment was applied to any of the specimens. The machined specimens were degreased in acetone and cleaned ultrasonically in alcohol before encapsulation in Type 304 SS capsules filled with research-grade He for irradiation in the Halden reactor.

SPECIMEN 1.D. BOTH ENDS - 57.20 (SEE NOTE 9)

28.60

6.35- 3.4 12.70 6.35 R (SEE NOTE 6) - 762

.822

12.70 "

6.35- Wl 4.85

19.10 -DA. DRILLTHRU GAGE 2 HOLES

SECTION (SEE NOTE 5)

W1 =3.15, W2= W1 + 0.05W2• W2

Fig. 1. Geometry of slow-strain-rate tensile specimens; all dimensions in mm

NUREG/CR-6687 -6-

3 Specimen Irradiation

A total of 96 SSRT and 24 CT specimens were encapsulated into six capsules, each

capsule containing 16 SSRT and 4 CT specimens. A fixed 0.5-mm gap was allowed between

the inner wall of the Type 304 SS capsule and specimen edges. The gap was filled with

research-grade He. The gap size of 0.5 mm was selected to maintain specimen temperature at

2880 C during irradiation in He. To prevent capsule wall creepdown and possible changes in

gap size, spacers in the form of Type 304 SS wires (0.5 mm diameter) were placed between the

specimens and the capsule inner wall. Type 304 SS filler bodies were inserted on both sides of

the SSRT specimen stack to avoid overheating the thin gauge section. Table 5 shows the

specimens categorized according to alloy, fluence level, and postirradiation test type. The six

capsules were irradiated in the Halden boiling heavy water reactor starting April 8, 1992. Fast

neutron (E > 1MeV) flux during the various irradiation cycles ranged from 1.80 x 1013 n cm- 2

s- 1 to 3.31 x 1013 n cm-2 s-1. Irradiation history of the six capsules is summarized in Table 6.

Table 5. Summary of specimen number per alloy, irradiation fluence,

and postirradiation test type

SSRT Uniaxial Constant J-R or Crack Specimen Test Load Test Growth Rate Test

ID higha mediuma lowa high medium low high medium low

C1 1 1 1

L2 1 1 1 1

C3 1 1 1 1 1 1

IA 1 1 1 i15 1 1 1 1 L6 1 1 L7 1 1

L8 1 1 1

C9 1 1 1

Clo 1 1 1 LII 1 1 1 C12 1 1 1 L13 1 1 1

L14 1 1 1

LIS 1 1

C16 1 1 1 1 1 L17 1 1

L18 1 1 1 1 1

C19 5 1 1 4 4 1 1 1

L20 5 1 1 4 4 1 1 1

C21 1 1 1 1 1 1

L22 1 1 1 1 1 L23 1 1 1

124 1 1 1

L25C3 3 L26C19 3 L27C21 2

aFluence level in 1021 n cm"2 (E > 1 MeV): high 2.5, medium 0.9, and low 0.3.

NUREG/ CR-6687-7-

Table 6. Summary of discharge fluence of model austenitic stainless steel alloys irradiated in Halden Boiling Heavy Water Reactor

Target Capsule Fluence Fluencea

ID Level (1021 n cm-2)

1 medium 1.0

4 low 0.4

5 high 2.5

6 high 2.5

7 medium 1.0

8 high 2.5 aFor neutron energy E > 1 MeV

Irradiation Cycles

IFA 530-3 to -6; D-07-004-2

IFA 530-3

IFA 530-4 to -6; D-07-004-1 to -3

IFA 530-4 to -5; D-07-004-1 to -3

IFA 530-4 and -6; D-07-004-1

IFA 530-5 to -6; D-07-004-1 to -3

Discharge Date

Nov. 96

Oct. 92

Nov. 99

Nov. 99

May 96

Nov. 99

Discharge Fluencea

(1021 n cm-)

0.9

0.3

2.0

2.0

0.9

2.0

NUREG/CR-6687 -8-

4 SSRT Test and SEM Fractography on Medium-Fluence Specimens

Slow-strain-rate tensile tests and fractographic analysis by SEM have been completed for

the 16 alloys that were irradiated to a fluence of =0.3 x 1021 n-cmn2 (E > 1 MeV) at =288°C in

an He environment in the Halden reactor. Tests were also completed on 23 medium-fluence

alloy specimens irradiated to -0.9 x 1021 n-cm-2 (E > 1 MeV). In addition to the irradiated

specimens, unirradiated control specimens were tested under the same conditions. All SSRT

tests were conducted at 2890C in simulated BWR water that contained =8 ppm DO.

Conductivity and pH of the water were kept at =0.07-0.10 and 6.3-6.8, respectively.

Strain rate was held constant at 1.65 x 10-7 sn-. Electrochemical potential was measured on

the effluent side at regular intervals.

Comprehensive results of SSRT testing and fractographic analysis of the unirradiated

control specimens and specimens irradiated to low fluence of =0.3 x 1021 n-cm- 2 (E > 1 MeV)

were reported previously. 1 9 For comparison with similar results obtained from specimens

irradiated to medium fluence of =0.9 x 1021 n-cm-2 (E > I MeV) given in this report, results

obtained for zero and low fluence are summarized in Tables 7-10. Feedwater chemistry (i.e.,

DO, ECP, conductivity, and pH) and results from SSRT testing (i.e., 0.2%-offset yield strength,

maximum strength, uniform plastic strain, and total plastic strain) are summarized in Tables

7 and 9, respectively, for nonirradiated control specimens and specimens irradiated to =0.3 x

1021 n-cm-2 (E > 1 MeV). Also shown in these tables are results of SEM fractographic analysis

of the failure mode (i.e., ductile, intergranular, and transgranular fracture surface morphology)

of the specimens. In Table 8, the results of SSRT and SEM fractographic analysis (percent

IGSCC, percent TGSCC, and combined percent IGSCC+TGSCC) are correlated with

compositional characteristics of the unirradiated specimens. Similar correlations for alloys

irradiated to =0.3 x 1021 n-cm-2 (E > 1 MeV) are given in Table 10.

4.1 Results of SSRT Tests and SEM Fractography

Results of SSRT tests and SEM fractography conducted on medium-fluence specimens

irradiated to -0.9 x 1021 ncm- 2 (E > 1 MeV) are summarized in Table 11. The results are

correlated with the compositional characteristics of the alloys in Table 12.

The results obtained from the SSRT tests and SEM fractography of the medium-fluence

specimens, i.e., yield strength, ultimate tensile strength, uniform elongation, total elongation,

percent TGSCC, percent IGSCC, and percent TGSCC+IGSCC, are plotted for each alloy in Figs.

2-8, respectively. When compared with the properties of the unirradiated and low-fluence

specimens, the effects of the increased fluence on yield strength, ultimate tensile strength,

uniform elongation, total elongation, and fracture behavior of the medium-fluence specimens

were significant. For example, the effects of fluence and test environment (i.e., air vs. water)

on load elongation behavior are shown in Fig. 9 for Heat C19, a commercially purchased heat

of Type 304 SS. For the commercial alloy, a significant effect of fluence on load-elongation

behavior in water is evident as fluence increases from 0 to -0.3 x 1021 ncm-2 and =0.9 x 1021

n c m- 2 . A similar effect of fluence was significantly more pronounced for most laboratory

alloys than for commercial alloys.

NUREG/CR-6687-9

Table 7. Results of SSR7h test and SEMfractography for nonirradiated control specimens of model austenitic stainless steel alloys.

Alloy and Feedwater Chemistry SSRT Parameters Fracture Behavior Specimen Oxygen Average Cond. Yield Max. Uniform Total TGSCC+

Ident. SSRT Conc. ECP at 25°C pH Stress Stress Elong. Elong. TGSCC IGSCC IGSCC No. No. (ppm) (mV SHE) (gS.cm-l) at 25°C (MPa) (MPa) (%) (%) (%) (%) (%)

L23-4 CHR-1 8.6 +228 0.07 6.65 332 480 15.6 17.0 15 0 15 L7-4 CHR-2 8.0 +217 0.07 7.37 195 370 2.5 5.2 20 0 20 L7-B1 CHR-7 tested in air 282 676 42.3 43.9 0 0 0 L14-4 CHR-3 8.6 +208 0.07 7,37 240 474 41.8 44.2 0 0 0 L17-4 CHR-4 7.5 +262 0.06 7.09 189 412 11.6 13.3 60 0 60 L17-B1 CHR-19 7.8 +166 0.08 6.71 184 447 30,1 31.2 8 0 8 L6-4 CHR-5 7,9 +256 0.08 6.85 227 515 43.0 44.5 0 0 0 L27-4 CHR-6 9.3 +247 0.08 6.96 298 483 20.6 22.9 0 0 0 L26-4 CHR-8 9.4 +223 0.07 6.65 184 506 38.2 40.2 0 0 0 L2-4 CHR-9 8.6 +292 0.06 6.55 193 348 6.6 7.8 57 0 57 L25-4 CHR-10 8.2 +239 0.06 6.42 184 458 25.5 27.0 0 0 0 L15-4 CHR-11 8.2 +195 0.06 6.32 218 512 36.7 37.9 0 0 0 L24-4 CHR-12 8.4 +200 0.07 6.20 352 461 10.4 12.3 10 0 10 Cl-15 CHR-13 8.1 +187 0.07 6.33 179 498 49.4 51.7 0 0 0 C19-B1 CHR-14 8.8 +179 0.08 6.29 178 501 47.4 49,2 0 0 0 C9-B1 CHR-15 8.5 +166 0.07 6.83 178 408 17.4 19.4 32 0 32 C12-Bl CHR-16 8.5 +124 0.07 6.18 182 511 46.0 47.6 0 0 0 CIO-BI CHR-17 9.2 +145 0.07 6.26 174 478 30.6 35.1 0 0 0 C21-9 CHR-18 9.2 +187 0.07 6.41 277 455 48.9 59.5 0 0 0 aTested at 289°C at a strain rate of 1.65 x 10-7 s71 in simulated BWR water containing -8 ppm DO.

NUREG/CR-6687 10-

Table 8. Compositionional characteristics (composition in wt.%) of nonirradiated control specimens of model austenitic stainless

steel alloys correlated with results of SSRT testa and SEMfractography (HP = high purity, CP = commercial purity).

Alloy 0 YS UTS LE TE TGSCC IGSCC TG+IGSCC

ID Ni Si P S Mn C N Cr Mo/Nb (wppm) Remarkb (MPa) (MPa) (%) (%) (%) (% M ()

L23 12.04 0.68 0.030 0.047 0.96 0.043 0.092 17.30 Nb 1.06 93 CP 348 332 480 15.6 17.0 15 0 15

L7 10.60 0.18 0.040 0.038 1.02 0.007 0.111 15.40 - 274 High N, 0; Low Si, C 195 370 2.5 5.2 20 0 20

L14 7.93 1.49 0.080 0.002 1.76 0.107 0.028 15.00 - 45 High Si, P, C; Low S 240 474 41.8 44.2 0 0 0

L17 8.00 0.66 0.090 0.009 0.48 0.061 0.078 15.30 - 90 High P; Low Cr, Mn, S 189 412 11.6 13.3 60 0 60

L17 8.00 0.66 0.090 0.009 0.48 0.061 0.078 15.30 - 90 High P; Low Cr, Mn, S 184 447 30.1 31.2 8 0 8

L6 10.00 1.90 0.020 0.005 1.13 0.096 0.087 17.10 - 58 High Si, C, Cr; Low S 227 515 43.0 44.5 0 0 0

L27 10.30 0.96 0.040 0.002 0.97 0.057 0.019 15.30 Mo 2.01 - CP 316 ; high B (0.030) 298 483 20.6 22.9 0 0 0

L26 8.09 0.79 0.004 0.002 0.91 0.070 0.089 17.20 - 80 Low P, S 184 506 38.2 40.2 0 0 0

L2 10.50 0.82 0.080 0.034 1.58 0.074 0.102 17.02 - 66 High P, S, Mn, N 193 348 6.6 7.8 57 0 57

L25 8.93 0.92 0.020 0.008 1.54 0.019 0.095 17.20 - 85 high B (0.010) 184 458 25.5 27.0 0 0 0

L15 8.00 1.82 0.010 0.013 1.07 0.020 0.085 17.80 - 110 High N; Low C 218 512 36.7 37.9 0 0 0

L24 12.30 0.03 0.007 0.005 0.48 0.031 0.002 16.90 Nb 1.72 - HP 348; Low Si, N 352 461 10.4 12.3 10 0 10

Cl 8.12 0.50 0.038 0.002 1.00 0.060 0.060 18.11 - - Low S, CP 304 179 498 49.4 51.7 0 0 0

C19 8.08 0.45 0.031 0.003 0.99 0.060 0.070 18.21 - - Low Si, S, CP 304 178 501 47.4 49.2 0 0 0

C9 8.75 0.39 0.013 0.013 1.72 0.062 0.065 18.48 - - Low Si, High Mn 178 408 17.4 19.4 32 0 32

C12 8.23 0.47 0.018 0.002 1.00 0.060 0.070 18.43 - - Low Si, S, P 182 511 46.0 47.6 0 0 0

C10 8.13 0.55 0.033 0.002 1.00 0.060 0.086 18.19 - - Low S, high N 174 478 30.6 35.1 0 0 0

C21 10.24 0.51 0.034 0.001 1.19 0.060 0.020 16.28 Mo 2.08 - CP 316; low B (0.001) 277 455 48.9 59.5 0 0 0

aTest at 289°C at a strain rate of 1.65 x 10-7 s-1 In simulated BWR water.

NUREG/CR-6687-11-

Table 9. Results of SSRYa test and SEMfractography for model austenitic stainless steels irradiated in helium at 289*C tofluence of0.3 x 1021 n-cm- 2 (E > 1 MeV).

Alloy and Feedwater Chemistry SSRT Parameters Fracture Behavior

Specimen Oxygen Average Cond. Yield Max. Uniform Total TGSCC + Ident. SSRT Conc. ECP at 25*C pH Stress Stress Elongation Elongation TGSCC IGSCC IGSCC

No. No. (ppm) (mV SHE) (iS-cm- 1) at 25°C (MPa) (MPa) (%) (%) (%) (N) (%)

Cl-i HR-I 8.3 +184 0.07 7.03 490 680 13.4 16.6 4 0 4 L5-1 HR-2 9.7 +208 0.07 6.89 513 539 29.5 32.7 2 2 4

L22-1 HR-3 8,0 +236 0.07 6.80 360 596 6.6 9.4 50 15 65 C3-1 HR-4 8,7 +161 0.07 6.68 338 491 27.7 31.6 5 0 5 C16-1 HR-5 8.3 +204 0.08 6.74 370 527 17.6 20.6 2 0 2 IA-1 HR-6 9.0 +202 0.08 6.70 367 542 19.7 22.3 46 0 46

L18-1 HR-7 9.0 +203 0.08 6.33 503 572 6.3 8.8 54 0 54 Cl0-1 HR-8 8.2 +174 0.07 6.35 523 640 17.4 18.9 6 0 6 C21-1 HR-9 8.1 +149 0.08 6.49 480 620 15.9 19.4 4 0 4 111-1 HR-10 9.0 +157 0.08 6.17 487 599 2.3 3.8 62 0 62

L13-1 HR-11 8.7 +164 0.08 6.17 248 461 22.1 24.8 8 0 8 L20-1 HR- 12 8.4 +174 0.07 6.20 454 552 2.9 5.1 32 2 34 C19-1 HR-13 9.5 +132 0.12 6.36 554 682 10.5 14.7 7 0 7 C9-1 HR-14 8.0 +192 0.11 6.30 522 607 13.4 14.6 24 0 24 C12-1 HR-15 9.0 +195 0.08 6.40 404 589 20.4 24.2 5 0 5 1,8-1 HR-16 9.0 +215 0.08 6.60 411 571 15.6 17.9 54 0 54

aTested at 289"C at a strain rate of 1.65 x 10- 7 s- 1 in simulated BWR water containing --8 ppm DO.

NUREG/CR-6687 12-

Table 10. Compositional characteristics (composition in wt.%) of model austenitic stainless steels irradiated tofluence of--0.3 x 1021 n-cm- 2 (E > 1 MeV) correlated with results of SSR7a test and SEM fractography (HP = high purity, CP =

commercial purity).

Mo/Nb Remark - Low S, CP 304

- High P, Cr; Low C

Mo 2.04 HP 316L, low Si, N - CP 304L, Low Si

- High Ni; Low Si, S - High Ni, Mn, C; Low N

- Low Si, N - Low S, CP 304

Mo 2.08 CP 316

- High P; Low Si, C, S, N - High Si; Low Mn, C, N

- HP 304L, Low Si, N

- Low Si, S - Low Si; High Mn

- Low Si, P, S - High Ni, P, Mn; Low Si, N

aTest at 2890C at a strain rate of 1.65 x I0-7 s-1 in simulated BWR water; DO =8 ppm.

YS UTS

(MPa) (MPa)

490 680

513 539

360 596

338 491

370 527

367 542

503 572

523 640

480 620

487 599

248 461

454 552

554 682

522 607

404 589

411 571

LE (%)

13.4

29.5

6.6

27.7

17.6

19.7

6.3

17.4

15.9

2.3

22.1

2.9

10.5

13.4

20.4

15.6

TE TGSCC IGSCC TG+IGSCC (M ) (0) (0) (0) 16.6 4 0 4

32.7 2 2 4 9.4 50 15 65

31.6 5 0 5

20.6 2 0 2 22.3 38 0 38

8.8 54 0 54

18.9 6 0 6

19.4 4 0 4

3.8 62 0 62 24.8 8 0 8

5.1 32 2 34

14.7 7 0 7 14.6 24 0 24

24.2 5 0 5 17.8 64 0 64

NUREG/CR-6687

Alloy

ID

C1

L5

L22

C3

C16

L4

L18

CI0

021

Lii

L13

L20

C19

C9

C12

L8

Ni Si

8.12 0.50

9.66 0.90

13.30 0.24

8.91 0.46

12.90 0.38

10.20 0.94

8.13 0.14

8.13 0.55

10.24 0.51

8.15 0.47

8.18 1.18

8.91 0.17

8.08 0.45

8.75 0.39

8.23 0.47

10.20 0.15

P S

0.038 0.002

0.113 0.028

0.015 0.004

0.019 0.004

0.014 0.002

0.031 0.010

0.016 0.033

0.033 0.002

0.034 0.001

0.097 0.009

0.027 0.022

0.010 0.004

0.031 0.003

0.013 0.013

0,018 0.002

0.093 0.010

Mn

1.00

0.47 0.40

1.81

1.66 1.75

1.13 1.00

1.19

1.02 0.36

0.41

0.99 1.72

1.00 1.85

C

0.060

0.006

0.003

0.016

0.020 0.110

0.080

0.060

0.060

0.014

0.026

0.002

0.060

0.062

0.060 0.041

N Cr

0.060 18.11

0.033 21.00

0.001 16.10

0.083 18.55

0.011 16.92

0.002 15.80

0.001 18.00

0.086 18.19

0.020 16.28

0.004 17.40

0.001 17.40

0.002 18.10

0.070 18.21

0.065 18.48

0.070 18.43

0.001 18.30

- 13-

Table 11. Results of SSR7a test and SEMftactography for model austenitic stainless steels irradiated in helium at

289°C to afluence of-0.9 x 1021 n.cn-2 (E> 1 MeV)

Feedwater Chemistry SSRT Parameter Fracture Behavior

Specimen Oxygen Average Cond. Yield Max. Uniform Total TGSCC +

Ident. SSRT Conc. ECP at 25°C pH Stress Stress Elong. Elong. TGSCC IGSCC IGSCC

No. No. (ppm) (mV SHE) (pS.cm- 1) at25°C (MPa) (MPa) (%) (%) (%) (%) (%)

L22-02 HR-17 8.0 +181 0.08 6.77 475 549 4.20 5.82 30 35 65

L11-02 HR-18 8.0 +191 0.08 6.55 820 856 0.43 1.65 50 14 64

L18-02 HR-19 8.0 +193 0.10 6.07 710 755 3.98 5.05 38 14 52

L20-02 HR-28 test in 289°C air 826 845 0.31 2.09 0 0 0

L20-05 HR-26 9.0 +182 0.09 6.32 670 743 0.37 1.03 0 0 0

L20-06 HR-27 8.0 +274 0.07 6.05 632 697 0.85 2.72 0 0 0

C9-02 HR-21 8.0 +240 0.07 6.47 651 679 1.42 2.50 62 22 84

L17-02 HR-22 8.0 +198 0.07 6.42 574 654 2.02 3.08 44 41 85

L7-02 HR-23 8.0 +215 0.07 6.03 553 561 0.24 2.44 38 54 92

C10-02 HR-24 7.0 +221 0.07 5.26 651 706 6.35 9.25 14 0 14

C3-02 HR-25 8.0 +240 0.07 6.34 632 668 16.72 19.74 9 4 13

C19-02 HR-30 test in 289°C air 888 894 6.41 10.21 1 0 1

C19-04 HR-31 8.0 +252 0.07 6.18 750 769 6.06 8.79 1 0 1

L6-02 HR-32 8.0 +250 0.07 6.40 493 546 2.45 3.77 8 27 35

L14-02 HR-33 8.0 +246 0.08 6.07 649 684 1.90 4.67 84 2 86

L13-02 HR-34 7.0 +222 0,09 6.85 602 624 1.67 4.95 55 2 57

L04-02 HR-35 7.0 +259 0.08 6.54 634 680 1.07 2.02 68 2 70

L05-02 HR-36 7.0 +243 0.07 6.85 665 725 3.07 4.57 3 5 8

C16-02 HR-37 7.0 +230 0.07 6.62 562 618 11.99 15.80 7 1 8

L8-02 HR-38 8.0 +242 0.07 6.57 838 838 0.12 3.12 15 22 37

C21-02 HR-39 8.0 +231 0.08 6.21 643 716 15.38 18.30 1 2 3

L2-02 HR-40 7.0 +239 0.07 7.11 839 849 0.88 1.56 38 4 42

L24-02 HR-41 8.0 +239 0.06 6.40 725 725 0.15 2.45 2 1 3

L23-02 HR-42 7.0 +237 0.08 6.60 787 818 0.38 1.24 3 24 27

C12-02 HR-43 7.0 +227 0.07 6.19 747 756 14.96 18.57 4 0 4

Cl-02 HR-44 8.0 +229 0.07 6.30 707 763 13.36 17.04 2 0 2

aTested at 289°C at a strain rate of 1.65 x 10-7 s-1 in simulated BWR water containing -8 ppm DO.

NUREG/CR-6687 14-

Table 12. Compositional characteristics (composition in wt.9%) of model austenitic stainless steels irradiated tofluence of -0.9 x 1021 n.cm-2 (E > 1 Me V) correlated with results of SSRTP test and SEMftactography

Specimen 0, B, YS UTS LE TE TGSCC IGSCC TG+IGSCCID Ni Si P S Mn C N

L22-02 13.30 0.024 0.015 0.004 0.40 0.003 0.001

Li 1-02

L18-02

L20-05

L20-06

C9-02

L17-02

L7-02

C10-02

C3-02

C19-04

L6-02

L14-02

L13-02

L4-02

L5-02

C16-02

L8-02

C21-02

L2-02

L24-02

8.15

8.13

8.91

8.91

8.75

8.00

10.60

8.13

8.91

8.08

10.00

7.93

8.18

10.20

9.66

12.90

10.20

10.24

10.50

12.30

0.47 0.097

0.14 0.016

0.017 0.010

0.017 0.010

0.39 0.013

0.66 0.090

0.18 0.040

0.55 0.033

0.46 0.019

0.45 0.031

1.90 0.020

1.49 0.080

1.18 0.027

0.94 0.031

0.90 0.113

0.38 0.014

0.15 0.093

0.51 0.034

0.82 0.080

0.03 0.007

0.009

0.033

0.004

0.004

0,013

0.009

0.038

0.002

0.004

0.003

0.005

0.002

0.022

0.010

0.028

0.002

0.010

0.001

0.034

0.005

1.02

1.13

0.41

0.41

1.72

0.48

1.02

1.00

1.81

0.99

1.13

1.76

0.36

1.75

0.47

1.66

1.85

1.19

1.58

0.48

0.014

0.080

0.002

0.002

0.062

0.061

0.007

0.060

0.016

0.060

0.096

0.107

0.026

0.110

0.006

0.020

0.041

0.060

0.074

0.031

0.004

0.001

0.002

0.002

0.065

0.078

0.111

0.086

0.083

0.070

0.087

0.028

0.001

0.002

0.033

0.011

0.001

0.020

0.102

0.002

Cr

16.10

17.40

18.00

18.10

18.10

18.48

15.30

15.40

18.19

18.55

18.21

17.10

15.00

17.40

15.80

21.00

16.92

18.30

16.28

17.02

16.90

L23-02 12.04 0.68 0.030 0.047 0.96 0.043 0.092 17.30

C12-02 8.23 0.47 0.018 0.002 1.00 0.060 0.070 18.43 C1-02 8.12 0.50 0.038 0.002 1.00 0.060 0.060 18.11

Mo, Nb Remarksb

Mo 2.04 HP 316L; low Si, N, S - high P; low Si, C, S, N

- low Si, N

O 0.0940 high 0; low Si, N; HP 304L

O 0.0940 highO; low Si, N; HP 304L - low Si; high Mn

O 0.0090 high P; low Cr, Mn, S

O 0.0274 high S, N, 0; low Si, C - CP 304; low S; high N - CP 304L; high Mn, N; low S

O 0.0200 CP 304; low S

O 0.0058 high Si; low S

O 0.0045 high Si, P, Mn; low Cr, S - high Si, S; Low Mn, C, N

- high Si, C; low N, Cr

- high Si, P, Cr; Low Mn, C

0.0157 high Ni; low P, S, C - high P, Mn; low Si, N

Mo 2.08 CP 316, low S

O 0.0066 high 0, P, S, N

Nb 1.72 HP 348L; low Si, P, S, C, N O 0.0129

Nb 1.06 CP 348, high S O 0.0093

- 304, low S, low P - 304. Iw R

aTest at 2896C at a strain rate of 1.65 x 10-7 s-1 in simulated BWR water; DO =8 ppm. bHP = high purity, CP = commercial purity.

(MPa) (MPa) (%) (%) (%) (%) (%) 475 549 4.20 5.82 30 35 65820

710

670

632

651

574

553

651

632

750

493

649

602

634

665

562

838

643

839

725

856

755

743

697

679

654

561

706

668

769

546

684

624

680

725

618

838

716

849

725

0.43

3.98

0.37

0.85

1.42

2.02

0.24

6.35

16.72

6.06

2.45

1.90

1.67

1.07

3.07

11.99

0.12

15.38

0.88

0.15

1.65

5.05

1.03

2.72

2.50

3.08

2.44

9.25

19.74

8.79

3.77

4.67

4.95

2.02

4.57

15.80

3.12

18.30

1.56

2.45

787 818 0.38 1.24

747 756 14.96 18.57

707 763 13.36 17.04

50 14 64

38 14 52

dendritic structure

dendritic structure

62 22 84

44 41 85

38 54 92 14 0 14

9 4 13

1 0 1

8 27 35

84 2 86

55 2 57

68 2 70

3 5 8

7 1 8

15 22 37

1 2 3

38 4 42

2 1 3

3 24 27

4 0 4 2 0 2

2 0 2

NUREG/CR-6687- is-

- 304 low Sv v • p .....

HP 348L-low Si 0.03, low N 0.002 ! CP 348-high S 0.047

HP 316L-low Cr 16.1, low Si 0.024, low N 0.001

0.3 x 102 'ncrm

2

U 0.9 x IO2 t

ncm"2

CP 316-low S 0.001

HP 304L-high 0 0.094, low Si 0.017, low N 0.002

L24

L23

L22

C21

L20

C19

L18

L17 016

L15 L14

L13

C12

Lli

C10 C9

CP 304-4Ow S 0.003

-low Si 0.14, low N 0.001

low Mn 0.48. high P 0.09

0, low P 0.014, low S 0.002

HP = high purity CP = commercial purity

SSRT in 289*C water, DO=,8 ppm

Si 1.49, high Mn 1.76, low S 0.002. low Cr 15.0

.18, low Mn 0.36, S 0.022. low N 0.001

CP 304-low S 0.002, low P 0.012

304L-high P 0.097. low N 0.004

CP 304-4Ow S 0.002, high N 0.086

304-ow Si 0.39

304-low Si 0.15. high P 0.093, low N 0.001

304L-low Cr 15.4. high 0 028, low Si 0,18, low C 0.007, high N 0.111

304-high Si 1.90, low S 0.005; UE 2.45 %

304L-high Si 0.90. high P 0.113. low C 0.006, high Cr 21.0

304-high Si 0.94. high C 0.11, low N 0.002, low Cr 15.80

CP 304L-high Mn 1.81, low S 0.004. high N 0.083

304-high P 0.080, high S 0.034. high N 0.10

- _CP 304-low S 0.002 "- -- - -- I I I. I I, I- I.

0 200 00 600 800 1000

Yield Strength in 289°C Water (MPa)

1 40U

Fig. 2. Effects offluence on yield strength measured in 289°C water containing =8 ppm DO.

HP 348L-low Si 0.03, low N 0.002

CP 348-high S 0.047

HP 316L-low Cr 16.1, low Si 0.024, low N 0.001

I CP 316-low S 0.001

HP 304L-high 0 0.094, low Si 0.017. low N 0.002

CP 304•-ow S 0.003

..... 304-low Si 0.14. low N 0.001

304-low Cr 15.3, low Mn 0.48, high P 0.09

304L-high Ni 12.90, low P 0.014. low S 0.002

ES 0.3xl10e' n c3

n"• 0 0.9 x 102' n cm"2

HP = high purity CP = commercial purity

SSRT in 289*C water, DO-8 ppm

304-high Si 1.49, high Mn 1.76. low S 0.002. low Cr 15.0

304L-high Si 1.18. low Mn 0.36. S 0.022. low N 0.001

OP 304-low S 0.002. low P 0.012

- ~ 304L-high

CP 304-low S 0.002. high

0.097, low N 0.004

"N 0.086

304-low Si 0.39

304L-low Cr

304-high Si 1

________________________________C

304-low Si 0.15, high P0.093. low N0.001

15A, high 0 0.28. low Si 0.18. low C 0.007, high N 0.111

.90, low S 0.005; UE 2.45 %

304L-high Si 0.90, high P 0.113. low C 0.006. high Cr 21.0

304-high Si 0.94, high C 0.11, low N 0.002, low Cr 15.80

P 304L-high Mn 1.81, low S 0.004. high N0.083

304-high P 0.080. high S 0.034. high N 0.10

. ,3014cwSo. 0 ,002 1

400 600 800 1000 1

Ultimate Tensile Strength in 2890C Water (MPa)

Iu . .. . . . . 400

Fig. 3. Effects offluence on maximum strength measured in 289°C water

containing --8 ppm DO.

NUREG/CR-6687

304

304-4ow Cr 15.3.

304L-high Ni 12.9

"(test invalid)

304-high

304L-high Si 1

0

L8

L7

L6

L5

L4

03 L2

C14

L24

L23

L22

C21

L20

C19

L18 L17

C16

L15

L14

L13

C12

L1i

C10

C9 L8

L7 L6

L5

nvalid)

a LU

a)

3,. 0

(test i

L4 C3 L2 C1

200

I

ý

• AAA d

-----

AAA

%

1

1200UA •A

1,00U

- 16-

H P .1.ýP48 l ow . 0 . '3,low ,,AlN lS. . .. . . .. . . . . .. . '.. .

* cP 348-Ngh S 0.047

5 HP 3161-low Cr 16.1, low Si 0.024. low N O.001

CP 316-low S 0.001

- ' HP 304L-high 0 0.094, low Si 0.017, low N 0.002

CP 304-low S 0.003

s 3044ow Si 0.14, low N 0.001

304-low Cr 15.3. low Mn 0.48, high P 0.09

0.3xl101ncm-'(E>1 MeV)

U 0.9x1I0'ncm'2 (E>1 MeV)

HP = high purity CP = commercial purity

SSRT in 2890C water, DO-8 ppm

L24

L23

L22

C21

L20

C19

L18 L17

C16

L15 L L14 SL13

C 012

>% L11

C10

C9 LB L7

L6

L5 L4

C3 L2

C1

304L-high Si 1.18, low Mn 0.36 ,S0.0S 2. low N 0.001

CvOP 304-low S0.002, low P 0.012

&33 304L-high P 0.097, low N 0.004

OP 304-low S 0.002, high N 0.086

o 304-40w SI 0.39

S................................. 304-low Si 0.15. high P 0.093, low N 0.001 I 30414ow Cr 15.4. high 0 0.28, low Si 0.18, low C 0.007, high N 0.111

304-high Si 1.90, low S 0.005; UE 2.45 % 304I.-high Si 0.90. high P 0.

304-high

- 304-highP0.080.high .034high N0.10

OP p04-low S0.002.

113, low C 0.006. high Cr 21.0

Si 0.94, high C 0.11, low N 0.002, low Cr 15.80'

CP 304L-high Mn 1.81. tow S 0.004. high N 0.083

0 5 10 15 20 25

Uniform Elongation (%)

30 35 40

Fig. 4. Effects offluence on uniform elongation measured in 289°C water containing =8 ppm DO.

L24 P 4 - Si 0.03. w N 0.002 . .. . .

L23 cPa S-aighSO.O4c M 0.3x102 1

ncm'2

(E>1MeV)

L22 H P316L-lowCr16.1, lowSi0.024,lowN0.001 U 0-.x1021 ncm' 2(E>l MeV)

C21 cP 316-low S 0.001 L20 N P 304L-high O0,094, low Si 0.017. lowNO 0.002 HP = h o m righ purity HP = high puri ity

L18o- ., lSSRT

in 289l0

Li 8304lowSi .14.lowN 0001water, DOtS ppm Li17 304-low Cr15.3, low Mn 0.48, high P0.09

016

304L-high Ni 12.90, low P00.014. low S 0.002

Li14 304-high Si 1.49, high Mn 1.76 , low S 0.002, low Cr 15.0 L 1 3 -304.,-hig

Si 1.16, low Mn 0.36, S 0,02.. low NO0.001 012

OP 304-low S 0.002. low P0.012 Li • 304L-high P 0.097. low N 0.004 010

OP P304-lwSO.0�, high N 0.06

L7 304-low Cr 15A, higho 0.28, lowS0.18 low0.007, high N 0.111 LB6 304-high Si 1.90, low S 0.001: Ul 2.45 % L5 -3041-high Si 0.90, high P 0.113 , low 0 0.006, high Cr 21.2

L4 ..304-hh Si 0.94, high C 0.11, low N 0,002, low Cr 15.80 L2 304-high P0.00, hih 90.034 high N 0.10

CP 304L-hgh Mn 1.61, low S 0.004, gh N 0.063 Cl

OP 304-olo 0 .002

L84 0-o i0.15. hihP003Io .001

0 5 10 15 20 25

Total Elongation (%)30 35 40

Fig. 5. Effects of fluence on total elongation measured in 289°C water containing =8 ppm DO.

NUREG/CR-6687

304L-high Ni 12.90, low P 0.014. low S 0.002

(ltog invalid)

S304-high Si 1.49, high Mn 1.76. low S 0.002, low Cr 15.0

Ca 0 0

IMM

- 17-

SSRT in 289°C TGSCC <2% means negligible TGSCC CP 348-high 5 0.047; UE 0.38% water, DO=8 ppm -,-.oo••,,,•.-..•, • •'••-...•., • • HP 316L-tow Cr 16.1. low Si 0.024,0lw NO0.001: US 4.2%;

- 4oS UHP

= high purity CP = commercial

purity•.••••HP 304L-high 0 0.094, tow Si 0.017. low N 0.002: US 0.85% US = uniform elongationSC 0-ow S 0.03 E .0_

• •....•.•..•. • .•.. • • .304-tow S510.14. tow NO0.001; US 3.98 % 304low Cr 15.3. tow Mn 0.48, high P 0.09: UE 2.02%

C304L-high Ni 12.90, tow P0.014, ow S 0.002; US 11.99% S(test invalid) 304-high Si 1.49, high Mn 1.76. low S 0.002, low Cr 15.0; UE 1.90 %

_ P304L-high

Si 1.18 tow Mo 0.36.S 0.022. low N0.001: US 1.67% -

L24

L23

L22

C21

L20

C19

L18

L17

C16

L15

Li4

L13

C12

Lii

C10

C9

L8

L7

L6

L5

L4

C3

L2

C1

CP 304-low S O.O0. high N 0.086; UE 6.35 %

304L-tow Cr 15.4,

304-high Si 1.90, low S 0.005; UE 2.45 %

l 304L-high Si 0.90. high P 0.113. low C 0.006, high Cr 21.0; UE 3.0

304-low Si 0.39; UE 1.42 %

304-tow Si 0.15. high P 0.093, low N 0.001; UE 0.12 %

high 0 0.28. low Si 0.18. low C 0.007, high N 0.111; UE 0.24%

07%

__ IIC_ 34LhghM_101_owS_.04_ig N008 E16 304-high Si 0.94, high C 0.11, low N 0.002, low Cr 15.80; UE 1.07 S CP 3041.-high Mn 1.81.1low S 0.004, high N 0.083; US 16.72 % 0.xioncm

304-high P 0.080, high S 0.034. high N 0.10; UE 0.88 %

: CP O•4-ow S. 0.002iUE l -6% .36... . . . . .. . . . . . , O .9 x 10 n c mrn 2 -

20 40 60

Percent TGSCC (%)

80 1 ,

Fig. 6. Effects of fluence on percent TGSCC measured in 2890 C water containing -=8 ppm DO.

I ,. , . . ; .. I . . .I I. . . .I I I I '4 P348L-tiowSi 0.ý tow N4 01102: USi 0.1 5% I

CP 348Ngh S 0.047; UE50.38%

HP 3161.-lOw Cr 16.1. tow Si 0.024 low N:0.001; UE 4.2%;:

r CP 316-loW S 0.001; UE 15.4%; HP = high puny SHP 304L-hihh 0.094, low Si 0.017. low N 0.002; UE 0.85% commercial purity

SCP 304-lo S 0.003; UE 6.06 U uniform %iongation

p 304-low Si 0.14. low N 0dA1; UE 3.98 % IGSCC <2% means negligible s * . . , 304-low Cr 115A low Mn 0.48, hjigh P 0.09; UE 2j%

i 304L-hiuh Ni 12.90. low *I 0.014. low • 0.002; UE 1 ".99 %

)le IGSCC

0;2%

. 1MeV)

.1 Mev)

UE 0.24 %

289°C ) -8 ppm

(test invalid)

i 3 0 4 -high Si: 1.49, high Mn: 1.76, low S 0fl02. low Cr 15.0; UE 1.90 %i

I 304L1-high Si 1.18. low Mn 0.6. S 0.022. low 0.001; UE 1.67

CP 304-low .S 0.002, low: P 0.012; UE 49 0 121 n cm "2

(E

a 304L-high P 0.097, low N .004; UE 0.43 0 nc SI 0.9 X 1 • n CM -2 (E

CP 304-1ow S 0.002, high: N 0.086; UE S.35 %

a 304-low Si O,39; UE 1.42 %

3044owSil.15, high P0.093 lowN 0.001: U5 0.12%

3b4L-ow Cr 15.4 ,high 0 0.28, low Si 0.18.low C 0.007 high N 0.111

304-high Si 1.90, tow:S 0.005; UE 2.45 %

S304L-hllgh Si 0.90, high P 0.113, lbw C 0 .000, hligh Cr 21.0; iE 3.07 %

l 304-high 9.1 0.94. high C 5.11, low N 0.002, low Cr 15480; UE 1.07 /o

j CP 304L-high Mn 1.81, low S 0.004: high N o.o0; US 16.72 %S

304-high b 0.066. high P d.080. highSO .igh N 0.10: .. 0.88% water, D

5cP 304-io, S 0.002; US;13.36 % i

0 10 20 30 40 50 60 70 80

Percent IGSCC (%)

Fig. 7. Effects of fluence on percent IGSCC measured in 2890 C water

containing =8 ppm DO.

NUREG/CR-6687

a 0 0 = 3,' 0

304L-high P 0.097. low N 0.004; UE 0.43 %

a

C

C

L24 L23

L22

C21

L20

C19

L18

L17

C16

L15

L14

L13

C12

Lii

C10

C9

L8

L7

L6

L5

L4

C3

L2

C1

0 002 low P 0 012 UE 14 96 %

F • l V l l . . . . t P I f l l t l l f I t l l l l l l ] ] I I I I I ] I I T

i

- 18-

I

Stress (MPa)

(p

Co 0

(�(b

CA�

� (b'

(-) i�., 0

(b

0

PO

+

+ o

M

0

Alloy Heat ID

-. t)C)..00 J0 D0 r- 0' r-rr 01 0 - r- (0 r-0 r- r - r- ~

0~~ 1 T vn r T ;'ý ~ ~ _ ... z. . .. ... Jo~ ' ... .. ... ...

pp

m C

0 0v

0 9 W -; - T.... . ..... . . . .... .

. . ........................ ... E....... .

~~0 Scar

0 p 0

M 0 0

S?~~a § a w

Cc c 0

0 e ~0 t.1 C) 11thA C) , n

0 '0 -x x 0z 0 ............... .... . . .OK.. . . . . .

0 a9r c 0c m

U)

0

0 ko

C-) 0'

00 I.

Results of the SEM fractography (see Figs. 6-8) show that when fluence increased from

=0.3 x 1021 n-cm- 2 to =0.9 x 1021 n-cm-2 , the percent TGSCC in some alloys decreased, and at

the same time, the percent IGSCC increased, e.g., Alloys L22, L18, Lii, and L8. The IGSCC fracture surfaces in these alloys were surrounded by TGSCC fracture surfaces This indicates that an IGSCC fracture surface emerges from a portion of a TGSCC fracture surface and that

for many alloys, high susceptibility to TGSCC is a precursor to susceptibility to IGSCC at a

higher fluence. However, the threshold fluence for the transition from TGSCC to IGSCC

appears to differ significantly from alloy to alloy.

4.2 Effect of Sulfur

Based on the information given in Tables 8, 10, and 12, we examined the relationship between the fracture properties and the compositional characteristics of the alloys to identify the alloying and impurity elements that significantly influence stress corrosion cracking behavior. In the unirradiated state or after irradiation to -0.3 x 1021 n-cm-2 , commercial and laboratory heats of Types 304, 304L, and 348 SS that contain relatively high concentrations of

S (>0.009 wt.% S, 15-18.50 wt.% Cr) exhibited significant susceptibility to TGSCC, whereas alloys that contain a relatively low concentration of S (<0.005 wt.% S) exhibited good resistance to TGSCC. These relationships are shown in Fig. 10.

100

90 SSRT at 289°C DO --8 ppm

80

70 alloys with S >0.009 wt.% 6-0 susceptible to TGSCC 60

0 50 black bar 0• unirradiated I-"40 specimen

4) Sgrey bar

30 0.3 x 102l' n cm 2

20alloys with S <0.005 wt.%

10 resistant to TGSCC

Cl C3C10C12916C19C21 L6 L14L24L26L27 L2 L7 L17C9 Alloy Heat ID

Fig. 10. Effect of S on susceptibility to TGSCC of two groups of

alloys in unirradiated state or after irradiation to =0.3 x 1021 n-cm- 2 (E > 1 Mev); one group contains low

concentrations of S (<0.005 wt.%) and the other contains relatively high concentrations of S (>0.009 wt. %).

NUREG/CR-6687 - 20 -

At =0.9 x 1021 n-cM'-2 , commercial heats of Types 304 and 316 SS that contain low concentrations of S (<0.004 wt.% S) exhibited high resistance to TGSCC and IGSCC (see Fig. 11) and high uniform and total elongation (see Fig. 12), whereas a commercial heat of Type 304 SS that contains a relatively high concentration of S (-0.013 wt.% S) exhibited poor resistance to TGSCC and IGSCC (Fig. 11) and low uniform and total elongation (Fig. 12). Susceptibilities of the high-S heat C19 to TGSCC were high in unirradiated state and after irradiation to low fluence (see Fig. 10).

100 "SSRT at 289'C

90 DO =8 ppm 0.9 x 10o n cm"2 304 SS

U 8 (E> 1MeV) a commercial alloy O 80 with S =0.013 wt.% U - susceptible to IASCC I- 70 Mn 0.99-1.72 wt.% + U O 60 Co 0"

-50- IG+TGSCC 30 O 40

030 all commercial alloys with , .S <0.004 wt.% resistant to IASCC C IGSCC *) 200 a3high Ni

10- 304 304 304 316

C1 C3 C10 C12 C16 C19 C21 C9

Alloy Heat ID

Fig. 11. Effect of S on susceptibility to TGSCC and IGSCC of commercial alloys measured after irradiation to =0.9 x 1021 n-cm-2 (E > I MeV); one Group of alloys contain low concentrations of S (<0.004 wt.%) and the other alloy contains relatively high concentrations of S (>0.013 wt. %).

The observations summarized in Figs. 10-12 indicate strongly that for commercial heats of Types 304 and 304L SS, a high concentration of S (>0.009 wt.%) is significantly detrimental and that a sufficiently low concentration of S (e.g., <0.004 wt.%) is a necessary condition for good resistance to IASCC.

From an earlier study on steels irradiated to =2.5 x 1021 n-cm-2 (E > 1 MeV) in a BWR, Kasahara et al. 16 reported that the effect of high concentrations of S on IASCC susceptibility was significant for one heat of Type 316L SS that contained 0.035 wt.% S, whereas for a similar heat of Type 316L SS that contained 0.001 wt.% S, susceptibility to IASCC was insignificant. However, a similar effect of the concentration of S was not observed for two heats of Type 304L SS. In other studies on steels irradiated to =0.67 x 1021 n-cm- 2 (E > 1 MeV), Tsukada and Miwa reported deleterious effects of high concentrations of S for a heat of

-21 - NUREG/CR-6687

Type 304L SS (0.032 wt.% S) 1 2 and a Ti-doped heat of Type 316 SS (0.037 wt.% S).2 0

Although these earlier investigations were conducted only on a limited number of heats, the results appear to be consistent with the present observation that an S concentration as low as =0.009 wt.% could be deleterious to the susceptibility of steel to TGSCC and IGSCC.

The exact mechanism of how a high concentration of S in Types 304, 304L, or 316 SS leads to higher susceptibilities to TGSCC and IGSCC, as observed in this study, is not clear. In a previous investigation, grain-boundary concentrations of S in BWR neutron absorber tubes and control blade sheath fabricated from several high- and commercial-purity heats of Types 304 and 304L SS and irradiated to =2.6 x 1021 n-cm- 2 (E > 1 MeV) were determined by Auger electron spectroscopy (AES).14 However, because of S contamination that occurs during the process of anodic charging of H into the specimen prior to analysis by AES (which is necessary to induce intergranular fracture in vacuum in an Auger electron microscope and hence reveal grain boundaries of the specimen), a good correlation could not be established between grain-boundary S concentration and susceptibility to IGSCC of the BWR neutron absorber tubes and control blade sheath.

30 Mn 0.99-1.72 wt.% SSRT at 2890C

DO -8 ppm

0.9 x 102 n cmr2 25 all commercial alloys with (E> 1 MeV)

o• S <0.004 wt.% ductility high

"U�nfr Total Elongation € 304LUniform

0 20- 304L Elongation •, 304

0) 304 high Ni 3 304 316 o 304\

J 15

0 I

"" 10 304 304

E a commercial alloy 0= with S -0.013 wt.% C 5ductility low

M Il~M 11M l lM lln111111,IM C1 C3 C10 C12 C16 C19 C21 C9

Alloy Heat ID

Fig. 12. Effect of S on uniform and total elongation of commercial alloys measured after irradiation to =0. 9 x 1021 n-cm-2 (E > 1 MeV); one group of alloys contain low concentrations of S (<0.004 wt.%); the other alloy contains a relatively high concentration of S (>0.013 wt.%}).

NUREG/CR-6687 - 22 -

Grain-boundary segregation of S in irradiated steels is also difficult to determine by

energy-dispersive spectroscopy (EDS) in a transmission electron or analytical electron microscope (TEM or AEM). Furthermore, several factors complicate the exact distribution of S

in irradiated steels, i.e., trapping of S in MnS and other sulfide precipitates, potential

dissolution of MnS and other sulfides in BWR-like water,2 1 and irradiation-induced

modification of the composition and morphology of MnS precipitates. 2 2 Therefore, it is not surprising that a good correlation could not be established between grain-boundary S

concentration and susceptibility of irradiated steels to IGSCC.

For unirradiated steels, thermally induced grain-boundary segregation of S has been

measured by AES by Andresen and Briant 2 3 for one heat of Type 304L and one heat of 316NG

SS that contained 0.030-0.037 wt.% S and were annealed at 400-700'C. Susceptibility of both materials to IGSCC was significant. The former material did not contain any Mn;

therefore, IGSCC in that material was attributed to grain-boundary segregation of S. The lower percent IGSCC observed for the latter material, which contained 1.1 wt.% Mn and 0.067 wt.% P, was attributed to lower grain-boundary segregation of S that may have occurred in the

material because P and S must compete for grain-boundary sites for segregation.

A similar model, based on irradiation-induced grain-boundary segregation of S, does not

appear to be consistent with the present observation that a higher concentration of S is

conducive to higher susceptibilities to TGSCC and IGSCC. Conclusive evidence for irradiationinduced grain-boundary segregation of S does not appear to have been established. As pointed out previously, the IGSCC fracture surface in high-S steels was often observed in the middle of and surrounded by TGSCC fracture surface. This finding appears to indicate that MnS, and possibly other soluble sulfides that may have more complex chemical compositions,

dissolves in crack tip water that initially advances via TGSCC and then releases into the crack-tip crevice water deleterious ions that are conducive to producing IGSCC. To identify

the exact processes and mechanisms, SSRT tests on higher fluence specimens and a systematic microstructural investigation will be helpful.

4.3 Effect of Chromium

At a fluence of =0.9 x 1021 n-cm- 2 , a laboratory alloy that contains an unusually high

concentration of Cr (--21 wt.%) exhibited excellent resistance to TGSCC and IGSCC, in spite of high S content (=0.028 wt.% S). In contrast to this, several heats containing relatively low

concentrations of Cr of <17.5 wt.% exhibited significant susceptibility to TGSCC and IGSCC (see Fig. 13). This observation indicates that Cr atoms in high concentration play a major role

in suppressing susceptibility to IASCC under BWR conditions.

4.4 Effect of Silicon

Yield strength of the model alloys, measured in BWR-like water at 289°C, was nearly

constant at -200 MPa in unirradiated state and was more or less independent of Si concentration (see Fig. 14). However, as fluence was increased to =0.3 x 1021 n-cm- 2 and -0.9 x 1021 n-cm- 2 , the degree of increase in yield strength was significantly lower for alloys that contain >0.9 wt.% Si than for alloys that contain <0.8 wt.% Si (Fig. 14). This finding indicates

NUREG/CR-6687- 23 -

that nature of irradiation-induced hardening centers and the degree of irradiation hardening are significantly influenced by alloy Si content. Silicon atoms in austenitic stainless steels occupy substitutional sites; therefore, they are likely to interact preferentially with irradiationinduced vacancy sites in the steel. This effect is likely to inhibit the formation of vacancy clusters or vacancy-impurity complexes, and is, therefore, conducive to a less significant irradiation-induced hardening. An effect similar to that of Si was, however, not observed for C and N.

100 alloys with Cr <17.5 wt.%

90 more susceptible to IG+TGSCC Laboratory Heats Only SSRT at 289°C

80 15.00 DO =8 ppm

0.9 x 1021 n a- 2

15.80 (E>1 MeV)

U) 60 17.40

I-+ 50 U Cr 17.02 u)

40

- 30

alloy with 21 wt.% Cr 0 20 resistant to IG+TGSCC

10 21.0

L2 L14 L13 L4 L5 Alloy Heat ID

Fig. 13. Effect of Cr on susceptibility to TGSCC and IGSCC of laboratory alloys measured after irradiation to --0.9 x 1021 n-cm- 2 (E > 1 MeV); alloys susceptible to IASCC contain relatively low concentrations of Cr (<17.4 wt. %); an alloy resistant to IASCC contains relatively high concentrations of Cr (-=21.0 wt. %).

Among laboratory heats of Types 304 and 304L SS, alloys that contain <0.67 wt.% Si exhibited significant susceptibility to IGSCC, whereas alloys with 0.8-1.5 wt.% Si exhibited negligible susceptibility to IGSCC. However, an alloy with =1.9 wt.% Si exhibited some degree of susceptibility to IGSCC (percent IGSCC = 27). These observations indicate that an Si concentration of --0.8 to =1.5 wt.% is beneficial in delaying the onset of or suppressing the susceptibility to IASCC. To determine if similar effects are evident at higher fluence levels would require testing of the high-fluence specimens that will be discharged after irradiation to =2.0 x 1021 n-cm- 2 (E > 1 MeV).

Consistent with the beneficial effects of Cr and Si and the deleterious effect of S discussed above, an alloy (L7) that contains an unusually low concentrations of Cr (15.4 wt.%) and Si (0.18 wt.%) and an unusually high concentration of S (0.038 wt.%) exhibited the worst susceptibility to IASCC at --0.9 x 1021 n-cm-2 (i.e., -=54% IGSCC, -=92% TGSCC+IGSCC). Alloy L7 also contained an unusually high concentration of 0 (0.027 wt.%). Considering deleterious

NUREG/CR-6687 - 24 -

I, ^

lOOO0-

800

600

400

0.

C

2.

V.0 0.2 0.4 0.6 0.8 1.0 1.2 1.4 Si Content of Alloy (wt.%)

1.6 1.8 2.0

Fig. 14. Effect of Si concentration on yield strength of Types 304 and 304L alloys measured in 289°C water before and after irradiation.

SSRT at 289°C DO -8 ppm

0.9x 102 1

ncm2, (E> 1 MeV)

alloys with Si <0.67 wt.% more susceptible to IGSCC

0.18

0.66

1 1 0.15

304 and 304L SSs Laboratory Heats

alloys with Si 0.8 -1.5 wt.% less susceptible to IGSCC

0.2 1.49 1.18 0.94 0.9 I I= = I I ,i I M

L11 L17 L18 L7 L8 L2 L14 L13 L4 L5

Alloy Heat ID

Fig. 15. Effect of Si on susceptibility to IGSCC of laboratory alloys of Types 304 and 304L SS measured after irradiation to =0.9 x 1021 n-cm- 2 (E > I MeV); alloys susceptible to IASCC contain relatively low concentrations of Si (<0.67 wt.%); alloys resistant to IASCC contain relatively high concentrations of Si (0.8-1.5 wt.%).

NUREG/CR-6687

qnnI i . --

304 and 304L SS SSRT at 289pC DO -- 8 ppm

*-0-9 x 10 2 ncm2

EB ES ES * S *

- 0.3 x e10 nncm'2

(U (U (UE9E

0 00c> 0 " CU 0 unirradiated

1 , , I , , I , , I , , I , .I ., . I . . 1, . , I

0

Io

a.

0 0.

- 25 -

n , ,A I f t I

I ý-VV

e') 0

1

effect of 0,14,15 the unusually high concentration of 0 is believed to be one of the important factors that led to the poor performance of the alloy. Alloy L7 exhibited significant susceptibility to TGSCC (-20% TGSCC) in water even in unirradiated state (see Table 7). These observations suggest that to avoid high susceptibility to IASCC, it is important to ensure sufficiently low concentrations of S and 0 and a sufficiently high concentration of Si.

4.5 Work-Hardening Capability

Although susceptibilities to TGSCC and IGSCC were insignificant and the fracture surface morphology was mostly ductile, some laboratory-fabricated alloys exhibited very low uniform elongation and poor work-hardening capability in water after irradiation to -0.9 x 1021 n-cm-2 , i.e., L20 (characterized by dendritic structure and very high 0 concentration), L8 (304 SS), L23 (348 SS), and L24 (348 SS). These alloys contained unusually high concentrations of 0 or unusually low concentrations of Si, or both. To better understand such behavior, which usually indicates severe susceptibility to dislocation channeling or other forms of localized deformation, a systematic microstructural investigation by transmission and scanning electron microscopies is desirable.

Laboratory alloys that contain unusually high concentrations of 0 exhibited either high susceptibility to IGSCC or poor work-hardening capability. Fracture surfaces of such alloys indicated that void coalescence and crack initiation are promoted at or near oxide inclusions that are in high density. Then, ductile tearing of remaining ridges occurs early, leading to low uniform and total elongations, indicating that careful control of 0 concentration, which is not specified in the ASTM Specifications for austenitic SSs, is a very important factor in minimizing susceptibility to IASCC.

NUREG/CR-6687 - 26 -

5 Conclusions

(1) At =0.3 x 1021 n-cm-2 , many alloys were susceptible to transgranular stress corrosion cracking (TGSCC), but only one alloy of Type 316L SS that contains a very low

concentration of Si (0.024 wt.%) was susceptible to intergranular stress corrosion cracking (IGSCC). Alloy-to-alloy variations in susceptibility to TGSCC were significant at -0.3 x 1021 n-cm-2.

(2) As fluence was increased from =0.3 x 1021 n-cm-2 to =0.9 x 1021 n-cm-2 , IGSCC fracture surfaces emerged in many other alloys, usually in the middle of and surrounded by TGSCC fracture surfaces. This finding indicates that for many alloys, high susceptibility to TGSCC is a precursor to susceptibility to IGSCC at a higher fluence.

(3) Alloy-to-alloy variations in susceptibility to TGSCC and IGSCC were significant at =0.9 x 1021 n-cm-2 . Susceptibility to TGSCC and IGSCC was influenced by more than one alloying and impurity element in a complex manner.

(4) In unirradiated state and at the relatively low fluence level of =0.3 x 1021 ncm-2, commercial and laboratory heats of Types 304, 304L, and 348 SS that contain relatively high concentrations of S (>0.009 wt.% S, 15-18.50 wt.% Cr) exhibited significant susceptibility to TGSCC, whereas alloys that contain a relatively low concentration of S (<0.005 wt.%) exhibited good resistance to TGSCC At =0.9 x 1021 n-cm-2 , commercial

heats of Types 304 and 316 SSs that contain low concentrations of S, exhibited high resistance to TGSCC and IGSCC and high levels of uniform and total elongation, whereas a commercial heat of Type 304 SS that contain a relatively high concentration of S

exhibited poor resistance to TGSCC and IGSCC and low levels of uniform and total elongation. These observations indicate that for commercial heats of Types 304 and 304L SS, a high concentration of S (>0.009 wt.%) is significantly detrimental and that a sufficiently low concentration of S is a necessary condition to ensure good resistance to IASCC.

(5) At -0.9 x 1021 n-cm- 2 , a laboratory alloy that contains a high concentration of Cr (=21 wt.%) exhibited excellent resistance to TGSCC and IGSCC in spite of high S content (=0.028 wt.% S), whereas heats with <17.5 wt.% Cr exhibited significant susceptibility to TGSCC plus IGSCC. This finding indicates that Cr atoms in high concentration play a major role in suppressing susceptibility to IASCC under BWR conditions.

(6) Yield strength of the model alloys, measured in BWR-like water at 289°C, was nearly

constant at -200 MPa in the unirradiated state and was more or less independent of Si concentration. However, as the alloys were irradiated to =-0.3 x 1021 n-cm-2 and -0.9 x 1021 n-cm-2 , the degree of increase in yield strength was significantly lower for alloys that contain >0.9 wt.% Si than for alloys that contain <0.8 wt.% Si. This observation indicates that the nature of irradiation-induced hardening centers and the degree of irradiation hardening are significantly influenced by alloy Si content. Similar influence of C and N was not observed.

(7) Among laboratory heats of Types 304 and 304L SS, alloys that contain <0.67 wt.% Si

exhibited significant susceptibility to IGSCC, whereas heats with 0.8-1.5 wt.% Si exhibited

NUREG/CR-6687- 27-

negligible susceptibility to IGSCC. However, an alloy with =1.9 wt.% Si exhibited some degree of susceptibility to IGSCC. These observations indicate that a Si content between =•0.8 and -1.5 wt.% is beneficial in delaying the onset of or suppressing the susceptibility to IASCC. Consistent with the effects of S, Cr, and Si, an alloy that contains unusually low concentrations of Cr and Si and unusually high concentrations of S and 0 exhibited the worst susceptibility to IASCC at =0.9 x 1021 n-cm-2 .

(8) Although susceptibility to TGSCC and IGSCC was insignificant and the fracture surface morphology was mostly ductile, some alloys exhibited very low uniform elongation and poor work-hardening capability in water after irradiation to =0.9 x 1021 n-cm-2 . Such alloys contained unusually high concentrations of 0 or unusually low concentrations of Si or both. To better understand such behavior, usually indicative of severe susceptibility to dislocation channeling or other forms of localized deformation, a systematic microstructural investigation by transmission and scanning electron microscopy would be needed.

(9) A few alloys that contain unusually high concentrations of 0 exhibited either high susceptibility to IGSCC or poor work hardening capability. This finding indicates that careful control of 0 concentration is a very important factor in minimizing susceptibility to IASCC.

NUREG/CR-6687 - 28 -

References

1. M. E. Indig, J. L. Nelson, and G. P. Wozadlo, "Investigation of Protection Potential against

IASCC," in Proc. 5th Intl. Symp. on Environmental Degradation of Materials in Nuclear

Power Systems - Water Reactors, D. Cubicciotti, E. P. Simonen, and R. Gold, eds.,

American Nuclear Society, La Grange Park, IL, 1992, pp. 941-947.

2. M. Kodama, S. Nishimura, J. Morisawa, S. Shima, S. Suzuki, and M. Yamamoto, "Effects

of Fluence and Dissolved Oxygen on IASCC in Austenitic Stainless Steels," ibid., pp. 948

954.

3. H. M. Chung, W. E. Ruther, J. E. Sanecki, A. G. Hins, and T. F. Kassner, "Effects of Water

Chemistry on Intergranular Cracking of Irradiated Austenitic Stainless Steels," in Proc.

7th Intl. Symp. on Environmental Degradation of Materials in Nuclear Power Systems

Water Reactors, G. Airey et al., eds., NACE International, Houston, 1995, pp. 1133-1143.

4. F. Garzarolli, P. Dewes, R. Hahn, and J. L. Nelson, "Deformability of High-Purity Stainless

Steels and Ni-Base Alloys in the Core of a PWR," in Proc. 6th Intl. Symp. on

Environmental Degradation of Materials in Nuclear Power Systems - Water Reactors, R. E.

Gold and E. P. Simonen, eds., The Minerals, Metals, and Materials Society, Warrendale,

PA, 1993, pp. 607-613.

5. H. Kanasaki, T. Okubo, I. Satoh, M. Koyama, T. R. Mager, and R. G. Lott, "Fatigue and

Stress Corrosion Cracking Behavior of Irradiated Stainless Steels in PWR Primary Water,"

in Proc. 5th Intl. Conf. on Nuclear Engineering, March 26-30, 1997, Nice, France.

6. A. J. Jacobs, G. P. Wozadlo, K. Nakata, T. Yoshida, and I. Masaoka, "Radiation Effects on

the Stress Corrosion and Other Selected Properties of Type-304 and Type-316 Stainless

Steels," in Proc. 3rd Intl. Symp. Environmental Degradation of Materials in Nuclear Power

Systems - Water Reactors, G. J. Theus and J. R. Weeks, eds., The Metallurgical Society,

Warrendale, PA, 1988, pp. 673-680.

7. K. Fukuya, K. Nakata, and A. Horie, "An IASCC Study Using High Energy Ion Irradiation,"

in Proc. 5th Intl. Symp. on Environmental Degradation of Materials in Nuclear Power

Systems - Water Reactors, D. Cubicciotti, E. P. Simonen, and R. Gold, eds., American

Nuclear Society, La Grange Park, IL, 1992, 814-820.

8. H. M. Chung, W. E. Ruther, J. E. Sanecki, A. G. Hins, and T. F. Kassner, "Stress

Corrosion Cracking Susceptibility of Irradiated Type 304 Stainless Steels," in Effects of

Radiation on Materials: 16th Int. Symp., ASTM STP 1175, A. S. Kumar, D. S. Gelles, R. K.

Nanstad, and T. A. Little, eds., American Society for Testing and Materials, Philadelphia,

1993, pp. 851-869.

9. H. M. Chung, W. E. Ruther, J. E. Sanecki, and T. F. Kassner, "Grain-Boundary

Microchemistry and Intergranular Cracking of Irradiated Austenitic Stainless Steels," in

Proc. 6th Intl. Symp. on Environmental Degradation of Materials in Nuclear Power

Systems - Water Reactors, R. E. Gold and E. P. Simonen, eds., The Minerals, Metals, and

Materials Society, Warrendale, PA, 1993, pp. 511-519.

NUREG/CR-6687- 29 -

10. J. M. Cookson, D. L. Damcott, G. S. Was, and P. L. Anderson, "The Role of Microchemical and Microstructural Effects in the IASCC of High-Purity Austenitic Stainless Steels," ibid., pp. 573-580.

11. M. Kodama, J. Morisawa, S. Nishimura, K. Asano, S. Shima, and K. Nakata, "Stress Corrosion Cracking and Intergranular Corrosion of Austenitic Stainless Steels Irradiated at 323 K," J. Nucl. Mater., 212-215 (1994) 1509.

12. T. Tsukada and Y. Miwa, "Stress Corrosion Cracking of Neutron Irradiated Stainless Steels," in Proc. 7th Intl. Symp. on Environmental Degradation of Materials in Nuclear Power Systems - Water Reactors, G. Airey et al., eds., NACE International, Houston, 1995, pp. 1009-1018.

13. F. Garzarolli, P. Dewes, R. Hahn, and J. L. Nelson, "In-Reactor Testing of IASCC Resistant Stainless Steels," ibid., 1055-1065.

14. H. M. Chung, W. E. Ruther, J. E. Sanecki, A. G. Hins, N. J. Zaluzec, and T. F. Kassner, "Irradiation-Assisted Stress Corrosion Cracking of Austenitic Stainless Steels: Recent Progress and New Approaches," J. Nucl. Mater., 239 (1996) 61.

15. J. M. Cookson, G. S. Was, and P. L. Anderson, "Oxide-Induced Initiation of Stress Corrosion Cracking in Irradiated Stainless Steels," Corrosion 54 (1998) 299.

16. S. Kasahara, K. Nakata, K. Fukuya, S. Shima, A. J. Jacobs, G. P. Wozadlo, and S. Suzuki, "The Effects of Minor Elements on IASCC Susceptibility in Austenitic Stainless Steels Irradiated with Neutrons," in Proc. 6th Intl. Symp. on Environmental Degradation of Materials in Nuclear Power Systems - Water Reactors, R. E. Gold and E. P. Simonen, eds., The Minerals, Metals, and Materials Society, Warrendale, PA, 1993, pp. 615-623.

17. A. Jenssen and L. G. Ljungberg, "Irradiation Assisted Stress Corrosion Cracking Postirradiation CERT Tests of Stainless Steels in a BWR Test Loop," in Proc. 7th Intl. Symp. on Environmental Degradation of Materials in Nuclear Power Systems - Water Reactors, G. Airey et al., eds., NACE International, Houston, 1995, 1043-1052.

18. G. Taguchi, in "Quality Engineering Using Robust Design," M. S. Phadke, ed., Prentice Hall, Engelwood Cliffs, NJ, 1989.

19. H. M. Chung, W. E. Ruther, and R. V. Strain, "Irradiation-Assisted Stress Corrosion Cracking of Model Austenitic Stainless Steels Irradiated in the Halden Reactor," NUREG/CR-5608, ANL-98-25, Argonne National Laboratory, March 1999.

20. T. Tsukada, Y. Miwa, and T. Kondo, "Effects of Minor Elements on IASCC of Type 316 Model Stainless Steels," in Proc. 7th Intl. Symp. on Environmental Degradation of Materials in Nuclear Power Systems - Water Reactors, G. Airey et al., eds., NACE International, Houston, 1995, pp. 1009-1018.

21. "Sulfide Inclusions in Steel," Proc. Intl. Symp., November 7-8, 1974, Port Chester, New York, J. J. de Barbadillo and E. Snape, eds., American Society of Metals, 1975.

NUREG/CR-6687 - 30 -

22. H. M. Chung, J. E. Sanecki, and F. A. Garner, "Radiation-Induced Instability of MnS

Precipitates and Its Possible Consequences on Irradiation-Induced Stress Corrosion

Cracking of Austenitic Stainless Steels," in Effects of Radiation on Materials: 18th Intl.

Symp., ASTM STP 1325, R. K. Nanstad, M. L. Hamilton, A. S. Kumar, and F. A. Garner,

eds., American Society for Testing and Material, 1999, pp. 647-658.

23. P. L. Andresen and C. L. Briant, "Role of S, P, and N Segregation on Intergranular

Environmental Cracking of Stainless Steels in High Temperature Water," in Proc. 3rd Intl. Symp. on Environmental Degradation of Materials in Nuclear Power Systems - Water

Reactors, G. J. Theus and J. R. Weeks, eds., The Metallurgical Society, Warrendale, PA, 1988, pp. 371-381.

NUREG/CR-6687-31 -

NRC FORM 335 U.S. NUCLEAR REGULATORY COMMISSION 1. REPORT NUMBER (2-89) (Assigned by NRC. Add Vol., Supp., Rev., NRCM 1102, andAddendurn Numbers, ff any.) 30,30 BIBLIOGRAPHIC DATA SHEET NUREG/CR-6687

(See kistrclions on Mhe reverse) ANL-00/21

2. TITLE AND SUBTITLE

Irradiation-Assisted Stress Corrosion Cracking of Model Austenitic Stainless Steel Alloys a DATE REPORT PUBLISHED

MONTH YEAR October 2000

4. FIN OR GRANT NUMBER

W6610 5. AUTHOR(S) 6. TYPE OF REPORT

H. M. Chung, W. E. Ruther, R. V. Strain, and W. J. Shack Technical; Topical 7. PERIOD COVERED (Inclusive Dates)

8. PERFORMING ORGANIZATION - NAME AND ADDRESS (If NRC, provide Division Office or Region, U.S. Nuclear Regulatory Commission, and mailing address; if contractor provide name and mailing address)

Argonne National Laboratory

9700 South Cass Avenue Argonne, IL 60439

9. SPONSORING ORGANIZATION - NAME AND ADDRESS (If NRC, type 'Same as above. if contractor, provide NRC Division, Office or Region, U.S. Nuclear Regulatoty Commission, and mailing address.)

Division of Engineering Technology

Office of Nuclear Regulatory Research U. S. Nuclear Regulatory Commission

Washington, DC 20555-0001

10. SUPPLEMENTARY NOTES

M. NcNeil, NRC Project Manager

11. ABSTRACT (200 words or less) This report summarizes work performed on irradiation-assisted stress corrosion cracking (IASCC) of model austenitic

stainless steels (SSs) that were irradiated in the Halden reactor in simulation of boiling water reactor core internal components. Slow-strain-rate tensile tests were conducted on 16 austenitic stainless steel alloys that were irradiated to a fluence of =3 x 1020 n cm- 2 (E > 1 MeV) and 23 alloys irradiated to =9 x 1020 n cm- 2 . Fractographic analysis by scanning electron microscopy was also conducted to determine susceptibility to IASCC as manifested by transgranular and intergranular fracture surface morphology. Susceptibility to IASCC was influenced by more than one alloying and impurity element in a complex manner. For commercial heats of Types 304 and 304L SS, a high concentration of S is detrimental and a sufficiently low concentration of S (i.e., <0.009 wt.0/o) is a necessary condition for good resistance to IASCC. An alloy containing a high concentration of =21 wt.% Cr exhibited excellent resistance to IASCC, despite a high S content, indicating that Cr atoms in high concentration play a major role in suppressing susceptibility to IASCC. After irradiation, the degree of increase in yield strength was significantly lower for alloys that contain >0.9 wt.% Si than for alloys that contain <0.8 wt.% Si, indicating that the nature of hardening centers and the degree of irradiation hardening are significantly influenced by the Si content of the alloy. Similar influence was not observed for C and N. Results also indicate that a Si content between =0.8 and =1.5 wt.% is beneficial in delaying the onset of or suppressing the susceptibility to IASCC. Some alloys exhibited low uniform elongation and poor work-hardening capability in water after irradiation to =0.9 x 1021 n-cm- 2, although the fracture surface morphology was mostly ductile. Such alloys contained unusually high concentrations of 0 or unusually low concentrations of Si or both.

12. KEY WORDS/DESCRIPTORS (List words or phrases t•at will assist researchers in locating tis report.) 13. AVAILABILITY STATEMENT

Boiling water reactor Unlimited

In-core water chemistry 14. SECURITY CLASSIFICATION

Core internal components (This Page)

Irradiation-assisted stress corrosion cracking Unclassified

Types 304 and 316 stainless steel (This Report)

Transgranular stress corrosion cracking Unclassified

Intergranular stress corrosion cracking 15. NUMBER OF PAGES Work-hardening capability

Tensile ductility 16. PRICE

Silicon, sulfur, oxygen, chromium

NRC FORM 335 (2-89)

Federal Recycling Program

NUREG/CR-6687 OCTOBER 2000IRRADIATION-ASSISTED STRESS CORROSION CRACKING OF MODEL AUSTENITIC STAINLESS STEEL ALLOYS

UNITED STATES NUCLEAR REGULATORY COMMISSION

WASHINGTON, D,C, 20555-0001

years