09/25/2007 - 09/26/2007 Meeting Presentation on Task 1 ... · Task 1: Evaluation of the Causes & Mechanisms of IASCC in BWRs - Crack Growth & Fracture Toughness of Irradiated Stainless

Work sponsored by the US Nuclear Regulatory Commission

Task 1: Evaluation of the Causes &Mechanisms of IASCC in BWRs -Crack Growth & Fracture Toughness ofIrradiated Stainless Steels

September 25-26, 2007

Nuclear Engineering Division

Argonne National Laboratory, Argonne, IL 60439

Investigators: Omesh Chopra, Gene Gruber, and Bill Shack

Experimental Effort: Ron Clark, Tom Galvin, and Loren Knoblich

2Work sponsored by theUS Nuclear Regulatory Commission

Objective

! Provide a better understanding of

– Threshold fluence above which the effects of neutron irradiation on

crack growth rates (CGRs) are significant

– Disposition curve for cyclic & SCC growth rates of irradiated SSs

– Fluence level above which benefit of HWC may be lost

! Significance of specimen size criteria

! Evaluate cyclic CGR data by using a superposition model

! Investigate the change in fracture toughness of austenitic SSs

under LWR irradiation conditions & temperatures

– Investigate effects of crack morphology (SCC IG vs. TG fatigue crack)

and BWR environment on fracture toughness

! Review the existing fracture toughness data in order to assess potential for

radiation embrittlement of reactor core internal components


Material

0.0200.0918.050.0680.0601.000.0080.0280.518.13304C19

0.0100.4418.230.0640.0131.840.0160.0270.539.05304LGG Top Shell

0.0130.5118.560.0840.0701.900.0070.0150.608.4530410285

0.722.5820.860.0520.0650.530.0120.0220.679.12CF-8M75

0.0140.3118.620.0670.0151.800.0080.0230.558.95304LGG Bottom Shell

0.0142.1016.270.0160.0601.230.0020.0350.6110.45316C21

0.0162.1816.910.0110.0291.650.0030.0260.4212.32316LC16

0.0140.1218.930.0740.0241.860.0030.0200.459.10304LC3

OMoCrNCMnSPSiNiSteelHeat ID

! CGR and/or fracture toughness J-R curve tests completed on

SA Types 304L, 304, 316L, & 316 SS irradiated up to !3 dpa;

sensitized 304 SS & HAZ of SAW & SMAW irradiated to !2.2 dpa; and

thermally aged CF-8M cast SS irradiated to !2.5 dpa

! Materials irradiated in the Halden heavy boiling water reactor in Norway;

SA SSs irradiated at !288°C & others at 297-300°C


Specimen Geometry

! Crack extension measured by DC potential drop method

! Current leads attached to the side of the specimen;

Potential leads attached across the notch

'C'

7.00

7.00

3.30

3.30

.794

CENTERED

3.00 DIA.

2 THRU HOLES

+.05- .00

15.00

14.00

6.50

'M'

A .02A

A .02

B

B .02

A .02

A .02

6.00

12.00

2.00

1.53 DIA

2 THRU HOLES

2.00

2.00

1.45

3.25

1.45

#56 (1.19) DIA. DRILL 3.25 DP.

#0-80 UNF-2B TAP 2.17 ±.06 DP. 2 HOLES.

XXX-X

SPECIMEN ID

C

C .02

C .02

.45 R

.45

DETAIL 'M'


Experimental Conditions

Temp: 289°C

DO: !350 ppb with N2 + 1% O2 cover gas

<30 ppb with 5% H2 cover gas

Flow: 15–25 mL/min

Conductivity: effluent 0.08 - 0.12 µS/cm

Cyclic Loading: load ratio 0.3-0.7

sawtooth waveform with 12 to 1000 s rise time

SCC: constant load with or w/o periodic partial unloading 1 or 2 h

Kmax: approximately constant by load shedding

K/size criterion: (W-a) "(2.5) (K/!effys)2 with effective yield stress defined as

!effys = (!ynonirr + !yirr)/2

J-R curve tests: constant extension rate of 0.026 mm/min

blunting line given by "a = J/(4!feff)


Environmental Enhancement of Growth Rates

! Under more rapid cycling loading typically used for precracking, crack growth is

dominated by mechanical fatigue

! For Kmax 15-18 MPa m1/2, environmental enhancement typically occurs at R "0.5 &

rise time "30 s; also fracture morphology changes from transgranular to intergranular

10-12

10-11

10-10

10-9

10-8

10-7

10-12 10-11 10-10 10-9 10-8 10-7

Best-fit 8 ppm DO

CG

Re

nv (

m/s

)

CGRair (m/s)

Austenitic SSs289°C

Precracking

Continuous Cycling


Enhanced Growth Rate for Irradiated Heat C3 of Type 304 SS

! Environmental enhancement observed after 170 h when load ratio & rise time

changed from 0.5 & 60 s to 0.7 & 300s

6.6

6.8

7.0

7.2

7.4

15

20

25

30

35

100 150 200 250 300 350 400

Cra

ck L

ength

(m

m)

Km

ax (

MP

a m

0.5

)

Time (h)

Type 304 SS (Heat C3)Test CGRI–07 (Spec. C3-B)

Fluence 0.9 x 1021 n/cm2

289°CHigh–Purity Water

CGR = 1.75 x 10–10 m/s

Kmax

= 20.1 MPa m0.5

R = 0.5, Rise Time 60 s

Kmax

Crack LengthDO !250 ppb

Steel ECP 190 mV (SHE)

DO <30 ppb

1.06 x 10–9 m/s

21.4 MPa m0.5

R = 1.0

1.04 x 10–9 m/s

23.5 MPa m0.5

R = 1.0

CGR = 6.38 x 10–10 m/s

Kmax

= 21.0 MPa m0.5

R = 0.7, Rise Time 300 s

Constant LoadUnload to R = 0.7 every 2 h


Data Analysis

Cyclic CGR data analyzed using the superposition model

CGRenv = CGRair + CGRcf + CGRscc

CGRair determined from correlation by James & Jones

CGRair = Css S(R) "K3.3/tr

S(R) = 1.0 R <0

S(R) = 1.0 + 1.8R 0 <R <0.79

S(R) = -43.35 + 587.97R 0.79 <R <1.0

Css = fn(T) and tr is the rise time

CGRcf based on expressions proposed by Shack & Kassner

CGRenv = CGRair + 4.5 x 10-5 (CGRair)0.5 !0.2 ppm DO

CGRenv = CGRair + 1.5 x 10-4 (CGRair)0.5 !8.0 ppm DO

CGRscc represented by correlation given in NUREG-0313

CGRscc = A (K)2.161

A = 2.1 X 10-13 for sensitized SS & !8 ppm DO

10-13

10-12

10-11

10-10

10-9

10-8

10-7

10-13 10-12 10-11 10-10 10-9 10-8 10-7

75-11TT

75-11TM

CG

Re

nv (

m/s

)

CGRair (m/s)

CF-8M Cast Austenitic SSIrradiated to 2.46 dpa289°C !300 ppb DO Water

Kmax

= 13 MPa m1/2

Irradiated SSModel 8 ppm DO

Specimen Number

Kmax

= 13 MPa m1/2

Irradiated SSModel 0.2 ppm DO


SCC Data for SSs Irradiated to 0.75-2.20 dpa

! Threshold fluence of 5 x 1020 n/cm2 (0.75 dpa) is inconsistent with experimental data

! At 0.75-2.20 dpa, CGRs are factor of 3-10 greater than those predicted by NUREG-0313

! CGRs of HAZ materials are generally greater than those of SA or sensitized SSs

! Benefit of HWC is observed at these fluence levels

10-12

10-11

10-10

10-9

10-8

10-7

5 10 15 20 25 30 35 40

304L 1.35 dpa316L 1.35 dpa316 1.35 dpa316NG 1.4-2.0 dpa304 Sensitized 2.16 dpa304L SAW HAZ 0.75 dpa304L SAW HAZ 2.16 dpa304 SMAW HAZ 0.75 dpa304 SMAW HAZ 2.16 dpa304 SMAW HAZ TT 0.75 dpaCF-8M Aged 2.46 dpa304 Sensitized 0.75 dpa

Experim

enta

l C

GR

(m

/s)

Stress Intensity K (MPa m1/2)

Material & Dose

NUREG-0313

Curve

6 x NUREG-0313Curve

Irradiated Stainless Steels 289°C

Open Symbols: NWC BWR Env.Closed Symbols: HWC BWR Env.

Data on 347 SS

from Halden


SCC Data for SSs Irradiated to <0.5 & 3.0-4.0 dpa

! At <0.5 dpa, CGRs comparable with values predicted by NUREG-0313

! At 3-4 dpa, benefit of HWC not observed for some heats at high K values

– tests considered invalid according to size criterion proposed by Andresen

10-12

10-11

10-10

10-9

10-8

10-7

5 10 15 20 25 30 35 40

304L 3.0 dpa

316 3.0 dpa

347 2.5-3.0 dpa

304 4.0 dpa

Experim

enta

l C

GR

(m

/s)

Stress Intensity K (MPa·m1/2)

Material & Dose

NUREG-0313

Curve

6 x NUREG-0313Curve



10-12

10-11

10-10

10-9

10-8

10-7

5 10 15 20 25 30 35 40

304L 0.45 dpa

316 0.45 dpa

Experim

enta

l C

GR

(m

/s)

Stress Intensity K (MPa·m1/2)

Material & Dose

NUREG-0313

Curve

6 x NUREG-0313Curve

Irradiated Stainless Steels NWC BWR Environment

289°C

347 SS data

from Halden &

304 SS data

from GE


Proposed K/Size Criteria for Irradiated Materials

! Two K/Size criteria have been proposed for irradiated materials which

generally show no strain hardening or actually show strain softening

(i.e., materials that deform by dislocation channeling)

– for moderate to highly irradiated materials (by Andresen)

!yeff = (!yirr+!ynonirr)/2

– for materials irradiated to very high fluences (by Anders)

!yeff = (!yirr+!ynonirr)/3

! However, basis for these criteria is not clear

! ANL tests have tried to evaluate the K/size criteria by

– consistency of results, &

– evidence of loss constraint in fractography


Benefit of Reduced DO Level (or ECP) on Growth Rates

! At Kmax !17.8 MPa m1/2, CGRs decreased a factor of !8 when

ECP decreased below -200 mV (DO from !500 ppb to <30 ppb)

! Rates increased back to old value when ECP increased above !100 mV

6.70

6.80

6.90

7.00

7.10

7.20

-600

-400

-200

0

200

400

50 100 150 200 250

ECP Pt

ECP SS

Cra

ck L

ength

(m

m)

EC

P (

mV

SH

E)

Time (h)

Type 316 SS (Heat C21)Test CGRI–26 (Spec. C21-C)


289°C, High–Purity Water

Crack Length

!500 ppb DO

!400 ppb DO

Kmax

= 17.9 MPa m1/2

17.6 MPa m1/2


Effect of Reduced DO Level on Growth Rates

! At the value allowed by !yeff = (!yirr+!ynonirr)/2, Kmax !24 MPa m1/2, no benefit

of reduced DO on CGRs even after ECP decreased below -200 mV

! In low-DO water, rates did not change significantly even when Kmax

decreased to !21 MPa m1/2

7.30

7.40

7.50

7.60

7.70

7.80

7.90

8.00

-600

-400

-200

0

200

400

300 350 400 450 500

ECP Pt

ECP SS

Cra

ck L

en

gth

(m

m)

EC

P (

mV

SH

E)

Time (h)



289°C, High–Purity Water

Crack Length

!400 ppb DO

24 MPa m1/2

!21 MPa m1/2

!25 MPa m1/2

!23 MPa m1/2


Specimen K/size Criterion

! There is no change in fracture plane, DO level was changed at 1.7 mm crack length;

fracture plane is straight & normal to stress axis

! If thickness or ligament criterion is exceeded, crack propagates away from the normal

plane at an angle of 45°

7.20

7.40

7.60

7.80

8.00

8.20

20

25

30

35

40

45

200 240 280 320 360 400 440 480 520

Cra

ck L

en

gth

(m

m)

Km

ax (

MP

a m

0.5

)

Time (h)



289°CConstant Load, periodic unloading

to R = 0.7 every 1 hKmax

Crack Length

CGR = 6.91 x 10–10 m/s

Kmax

= 27.5 MPa m0.5

Steel ECP -294 mV (SHE)

5.07 x 10–10 m/s

23.7 MPa m0.5

Steel ECP 165 – 10 mV (SHE)

!400 ppb DO!80 ppb DO

!20 ppb DO

! Expected decrease in CGR not

observed when DO decreased

from 400 to 20 ppb.

! If applied load had exceeded

the value allowed by K/size

criterion, then CGRs should

have increased in high-DO

water

! Loading conditions seem to

have had no effect until the DO

level was decreased


Specimen K/size Criterion (Contd.)

! No change in fracture morphology, complete intergranular fracture during SCC test.

DO level was decreased at 1.7 mm crack length

Location D


Specimen K/size Criterion (Contd.)

! Arguments against the proposed K/size criterion:

– strain softening in irradiated materials is rarely more than 10-15%

– in most plastic zones, the plastic strains are so low that the material never

passes the max tensile stress

– FEA indicate difference between strain distributions ahead of a advancing

crack, in a strain-hardening vs. strain-softening material, is marginal

! Adequacy of proposed K/size criterion for irradiated SSs needs to be examined

0

200

400

600

800

1000

0 1 2 3 4 5

325°C air

289°C air

Str

ess (

MP

a)

Strain (%)

Type 304 SS (Heat C19)Irradiated to 3.0 dpa

Strain rate 1.65 x 10-7

s-1

Rapid stress reduction is

due to necking


SCC Data for SSs Irradiated to !13 dpa

! CGRs show strong dependence on K at less than 15 MPa m1/2

! At K >15MPa m1/2, CGRs may be factor of 30 higher than NUREG-0313 curve

! Beneficial effect of low corrosion potential not observed at 13 dpa

10-12

10-11

10-10

10-9

10-8

10-7

5 10 15 20 25 30 35 40

304L 12.9 dpa

304L 13.1 dpa

Experim

enta

l C

GR

(m

/s)

Stress Intensity K (MPa m1/2)

Material & Dose

NUREG-0313

Curve

6 x NUREG-0313Curve



Data from

Studsvik & Halden


Fatigue CGR Data for SS Weld HAZs

! CGRs of nonirradiated weld HAZ are consistent with Shack/Kassner model

! Irradiation up to 2.2 dpa has only marginal effect on CGRs in air

10-12

10-11

10-10

10-9

10-8

10-7

10-12 10-11 10-10 10-9 10-8 10-7

85-3A-TT

85-YA

GG3B-A-TT

GG5B-A

CG

Re

nv (

m/s

)

CGRair (m/s)

Non Irradiated SS Weld HAZ

300–500 ppb Dissolved Oxygen

Symbols with +: K/size criterion not satisfied

Specimen NumberType 304 SMA Weld HAZ

Kmax

= 17 MPa m1/2

Nonirradiated SS

Model 0.2 ppm DO

Type 304L SA Weld HAZ

Kmax

= 17 MPa m1/2

Nonirradiated SSModel 8 ppm DO

10-12

10-11

10-10

10-9

10-8

10-7

10-12 10-11 10-10 10-9 10-8 10-7

Type 304 SS SMA Weld HAZ

Type 304L SS SA Weld HAZ

CG

Renv (

m/s

)

CGRair (m/s)

SS Weld HAZIrradiated to 2.16 dpa

Kmax

= 13 MPa m1/2

Irradiated SS

Model 8 ppm DO

Tested at 289°C in Open Symbols: !300 ppb DO Water

Closed Symbols: Air


Fatigue CGR Data for Irradiated SSs in NWC & HWC Water

! At >0.5 dpa, cyclic CGRs in NWC represented by CGRscc for irradiated steel

(i.e., 6 x NUREG-0313 rates) & Shack/Kassner model for 8 ppm DO

! At <0.5 dpa in NWC & irradiated SSs in HWC, cyclic CGRs represented by

CGRscc given in NUREG-0313 & Shack/Kassner model for 0.2 ppm DO

10-12

10-11

10-10

10-9

10-8

10-7

10-12 10-11 10-10 10-9 10-8 10-7

304L 0.45 dpa

304L 1.35 dpa

304L 3.00 dpa

316L 1.35 dpa

316 1.35 dpa

316 0.45 dpa

316 3.00 dpa

CG

Renv (

m/s

)

CGRair (m/s)

Austenitic SSs, 289°C!300 ppb DO

Kmax

= 14 MPa m1/2

Nonirradiated SSModel 0.2 ppm DO

Kmax

= 16 MPa m1/2

Irradiated SS Model 8 ppm DO

10-12

10-11

10-10

10-9

10-8

10-7

10-12 10-11 10-10 10-9 10-8 10-7

304 SS 1.35 dpa

316L SS 1.35 dpa

CG

Re

nv (

m/s

)

CGRair (m/s)

Austenitic SSs, 289°C<10 ppb Dissolved Oxygen

Nonirradiated SSModel 0.2 ppm DO

CGRair +4.5x10-5CGRair0.5

No SCC in low–DO Water


Fatigue CGR Data for Irradiated SS HAZs in NWC Water

! Cyclic CGR data for weld HAZ materials are similar to those for SA SSs;

CGRscc given by 6 x NUREG-0313 growth rates &

CGRcf given by Shack/Kassner model for 8 ppm DO

10-12

10-11

10-10

10-9

10-8

10-7

10-12 10-11 10-10 10-9 10-8 10-7

GG5T-A

GG5T-B

GG6T-A

CG

Re

nv (

m/s

)

CGRair

(m/s)

Grand Gulf Core Shroud

Type 304 L SA Weld HAZ289°C, !500 ppb DO

Kmax

= 13 MPa m1/2


Open Symbols: 0.75 dpaClosed Symbols: 2.16 dpa

Specimen Number

10-12

10-11

10-10

10-9

10-8

10-7

10-12 10-11 10-10 10-9 10-8 10-7

85-1A-TT

85-7A

85-XA

85-3TT

CG

Renv (

m/s

)

CGRair

(m/s)

Type 304 SS (Heat 10285)

289°C, !500 ppb DO

Kmax

= 13 MPa m1/2


Open Symbols: 0.75 dpaClosed Symbols: 2.16 dpa

SMA Weld HAZ

Sensitized


Fatigue CGR Data for Irradiated Cast SS in NWC Water

! Under similar loading & environmental conditions, cyclic CGRs for CF-8M

cast SS appear to be lower than those for wrought SSs & HAZ materials

10-13

10-12

10-11

10-10

10-9

10-8

10-7

10-13 10-12 10-11 10-10 10-9 10-8 10-7

75-11TT

75-11TM

CG

Re

nv (

m/s

)

CGRair (m/s)

CF-8M Cast Austenitic SSIrradiated to 2.46 dpa

289°C !300 ppb DO Water

Kmax

= 13 MPa m1/2


Specimen Number

Kmax

= 13 MPa m1/2

Irradiated SSModel 0.2 ppm DO


Summary

! Threshold fluence for irradiation effects to be significant: !3x1020 n/cm2 (!0.45 dpa);

below threshold, experimental CGRs are comparable to those of NUREG-0313

! Disposition curve for CGRs of SSs irradiated to 4 dpa: 6-8 x NUREG-0313 curve

! CGRs of SSs irradiated to 13 dpa show strong dependence of K and

are a factor of 30 higher than the NUREG-0313 curve

! Fluence level above which benefit of HWC is not observed:

limited data suggest, for some SSs it may be as low as !2x1021 n/cm2 (!3.0 dpa),

additional data needed to establish the threshold fluence above which irradiation

effects are significant in low-DO HWC BWR or PWR environments

! Adequacy of the proposed K/size criterion for irradiated SSs needs to be examined

! Cyclic CGRs of irradiated SSs can be represented by a superposition model


Fracture Toughness of Irradiated SSs - Background

! Much of the existing fracture toughness

data has been obtained in fast reactors

at temperatures above 350°C

! Exposure to neutron irradiation for

extended periods

- alters the microstructure

- increases the yield strength

- reduces ductility

- reduces resistance to fracture

! Fracture resistance decreases

substantially for 1-10 dpa, no further

decrease above saturation at !10 dpa

! Fracture toughness data

needed at LWR temperatures

0

200

400

600

800

1000

1200

0 5 10 15 20 25

Michel & Gray, 1987

Van Osch et al., 1997

Dufresne et al., 1979

Mills et al., 1985

Mills, 1988

Bernard & Verzeletti, 1985

Picker et al., 1983

Ould et al., 1988

JIc

(kJ/m

2)

Neutron Exposure (dpa)

Types 304 & 316 SSIrradiation Temp: 350 - 450°CTest Temp: 350 - 427°C


Effect of Irradiation on Fracture Toughness of SSs

! Neutron irradiation decreases the fracture toughness of SSs

! For the same irradiation level,

- toughness of cast CF-8M SS is lower than that of weld HAZ material, and

- toughness of HAZ material is lower than that of sensitized material

0

100

200

300

400

500

600

700

0 0.5 1 1.5 2 2.5 3

0.45 dpa

1.35 dpa

3.00 dpa

J (

kJ/m

2)

Crack Extension (mm)

Type 304 SS (heat C19) 289°C Air

J = 575!a0.17

JIC

= 503 kJ/m2

J = 438!a0.33

JIC

= 308 kJ/m2

J = 265!a0.29

JIC

= 184 kJ/m2

0

100

200

300

400

500

600

0 0.5 1 1.5 2 2.5 3

J (

kJ/m

2)


Austenitic Stainless Steels289°C BWR Water

304 Sensitized 10.5 h @ 600°C

CF-8M 2.46 dpa

304L SA Weld HAZ 2.16 dpa

304 SMA Weld HAZ 2.16 dpa

2.16 dpa0.75 dpa


Effect of Environment on Toughness of Irradiated Weld HAZs

! Limited data indicate that fracture toughness of irradiated Type 304L SAW

HAZ is approximately the same in air and water environments

! Complete J-R curve for Type 304 SMAW HAZ not obtained in air;

large crack extension occurred at same J value both in air and water

environments

0

50

100

150

200

250

300

350

400

0 0.5 1 1.5 2 2.5 3 3.5

GG6T-B in Air

GG6T-A in BWR Water

J (

kJ/m

2)


Type 304L SS, SAW HAZ


289°C Air

Estimated Effective Flow Stress: 502 MPa

0

50

100

150

200

250

300

350

400

0 0.5 1 1.5 2 2.5 3 3.5

Spec. 85-XB in Air

Spec. 85-XA in BWR Water

J (

kJ/m

2)


TYPE 304 SS, SMAW HAZ


289°C

Estimated Effective Flow Stress: 528 MPa

J = 219!a0.43

JIC = 128 kJ/m2


Effect of Environment on Toughness of Irradiated Sensitized SSs

! Although material tested in air was sensitized for longer time than the material

tested in water, toughness is slightly higher in air

0

100

200

300

400

500

600

0 0.5 1 1.5 2 2.5 3

10.5 h at 600°C

24.0 h at 600°C

J (

kJ/m

2)


Type 304 SS (Heat 10285


290°C

J = 316!a0.45

JIC

= 176 kJ/m2

Open Symbols: WaterClosed Symbols: Air

Heat Treatment

J = 376!a0.38

JIC

= 238 kJ/m2


Effect of Environment on Toughness of Irradiated Cast SSs

! Tests on thermally aged and irradiated cast SS in air were not conducted

! Both tests in water show large load drops and 0.5-1.0 mm crack extension;

such behavior is typically not observed during tests in air

0

100

200

300

400

500

600

700

0 0.5 1 1.5 2 2.5 3

J (

kJ/m

2)


CF-8M Cast SS (28% Ferrite)Aged 10,000 h at 400°C289°C Air

Open Symbols: 1/4-T CTClosed Symbols: 1-T CT

Irradiated to 2.46 dpaTested in BWR Water

0.0

1.0

2.0

3.0

4.0

5.0

6.0

0 0.5 1 1.5 2 2.5 3

75-11TM

75-11TT

Lo

ad

(kN

)

Displacement (mm)

CF-8M Cast SS (Heat 75, 28% ferrite) Aged 10,000 h at 400°C & Irradiated


289°C High-Purity Water

Specimen No.


Change in JIc of Austenitic SSs with Neutron Dose

! Data for BWR-irradiated materials within the scatter band for fast reactor

! JIc can decrease to !15 kJ/m2 (KIc = 54 MPa m1/2) at 3-5 dpa

! Data obtained in BWR water are generally lower than those obtained in air

0

100

200

300

400

500

600

0 5 10 15

JAPEIC BB

Japeic CT

JAPEIC SR

GE CT

304

316L

304 Sensi 10.5 h

304 SMA Weld HAZ

CF-8M

304L SA Weld HAZ

304 Sensi 24 hJIc

(kJ/m

2)


Austenitic SSs

ANL Heats

835 kJ/cm2

Closed Symbols: BWR WaterOpen Symbols: Air


Change in Coefficient C of Power-Law J-R Curve forCast SSs & Weld Metals with Neutron Dose

! For fluence less than 5 dpa, existing data can be bounded by a power-law

J-R curve with coeff. C expressed by the curve and exponent n = 0.37

0

150

300

450

600

750

900

0.01 0.1 1 10 100

308, Fast, 100-427, 125-427316, Fast, 370, 370316L, Fast, 90-250, 100-250CF-8, Fast, 400-427, 427CF-8, BOR-60, 325, 25CF-8M, Halden, 288, 289

Pow

er

Law

Consta

nt C

(kJ/m

2)


C = 20 + 205 exp(-0.65·dpa)

Cast Austenitic SSs & Welds


Change in Coefficient C of Power-Law J-R Curve forAustenitic SSs with Neutron Dose

! The power-law J-R curve expression yields a bounding C value of

225 kJ/m2 (1285 in-lb/in2) for materials irradiated less than 0.5 dpa &

28 kJ/m2 (160 in-lb/in2) for materials irradiated to about 5 dpa

0

150

300

450

600

750

900

0.01 0.1 1 10 100

304 BWR, 288, 289304 Fast, 100-427, 125-427316 Fast, 300-427, 300-427348 Fast, 385-413, 427304L BWR316CW Fast, 400-427, 205-427316L BWR, 288, 289316H Fast, 350, 350304 HAZ BWR, 288, 289304 HAZ Fast, 125-155, 125316H HAZ Fast, 350, 350304 BWR Sensi 10.5 h, 296, 289304 BWR Sensi 24 h, 296, 289304L HAZ BWR, 296, 289

Po

we

r L

aw

Co

nsta

nt

C (

kJ/m

2)


C = 20 + 205 exp(-0.65·dpa)

Austenitic Stainless Steels

Data with X tested in BWR water


Experimental Values of J-integral at 2.5 mm CrackExtension for SSs as a Function of Neutron Exposure

0

200

400

600

800

1000

1200

0.01 0.1 1 10 100

308, Fast, 100-427, 125-427316, Fast, 370, 370316L, Fast, 90-250, 100-250CF-8, Fast, 400-427, 427CF-8M, Halden, 288, 289

J a

t 2

.5 m

m C

rack E

xte

nsio

n (

kJ/m

2)


Austenitic Stainless SteelsWeld Metals

J = 255 kJ/m2

! EPRI TR-106092 proposed threshold value of J2.5 = 255 kJ/m2 for

potentially significant reduction in toughness of thermally aged cast SSs

! For SSs irradiated up to 0.3 dpa (2 x 1020 n/cm2), J2.5 is above 255 kJ/m2


Summary

! Neutron irradiation decreases fracture toughness of austenitic SSs

! For irradiated SSs, toughness of cast SS is lower than that of weld HAZ material,

and toughness of HAZ is lower than that of SA or sensitized SS

! J-R curve data for irradiated 304L SAW HAZ in water are comparable to those in air

! However, results for irradiated sensitized SS and cast SS suggest a possible effect

of water environment; additional tests are needed to verify these results

! Existing data indicate little or no change in toughness below 0.5 dpa, &

rapid decrease between 1 and 5 dpa to reach a saturation value

! A fracture toughness trend curve that bounds the existing data has been defined in

terms of JIc vs. neutron dose or coeff. C of power-law J-R curve vs. neutron dose

! For fluence less than 5 dpa, existing data can be bounded by a power-law J-R curve

with coeff. C expressed by C = 20 + 205 exp(-0.65 dpa) and exponent n = 0.37

09/25/2007 - 09/26/2007 Meeting Presentation on Task 1 ... · Task 1: Evaluation of the Causes & Mechanisms of IASCC in BWRs - Crack Growth & Fracture Toughness of Irradiated Stainless

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