09/25/2007 - 09/26/2007 Meeting Presentation on Task 1 ... · Task 1: Evaluation of the Causes &...
Transcript of 09/25/2007 - 09/26/2007 Meeting Presentation on Task 1 ... · Task 1: Evaluation of the Causes &...
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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
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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
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3Work sponsored by theUS Nuclear Regulatory Commission
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
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4Work sponsored by theUS Nuclear Regulatory Commission
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'
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5Work sponsored by theUS Nuclear Regulatory Commission
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)
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6Work sponsored by theUS Nuclear Regulatory Commission
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
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7Work sponsored by theUS Nuclear Regulatory Commission
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
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8Work sponsored by theUS Nuclear Regulatory Commission
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
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9Work sponsored by theUS Nuclear Regulatory Commission
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
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10Work sponsored by theUS Nuclear Regulatory Commission
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
Irradiated Stainless Steels 289°C
Open Symbols: NWC BWR Env.Closed Symbols: HWC BWR Env.
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
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11Work sponsored by theUS Nuclear Regulatory Commission
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
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12Work sponsored by theUS Nuclear Regulatory Commission
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)
Fluence 2.0 x 1021 n/cm2
289°C, High–Purity Water
Crack Length
!500 ppb DO
!400 ppb DO
Kmax
= 17.9 MPa m1/2
17.6 MPa m1/2
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13Work sponsored by theUS Nuclear Regulatory Commission
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)
Type 316 SS (Heat C21)Test CGRI–26 (Spec. C21-C)
Fluence 2.0 x 1021 n/cm2
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
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14Work sponsored by theUS Nuclear Regulatory Commission
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)
Type 304 SS (Heat C3)Test CGRI–08 (Spec. C3-C)
Fluence 2.0 x 1021 n/cm2
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
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15Work sponsored by theUS Nuclear Regulatory Commission
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
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16Work sponsored by theUS Nuclear Regulatory Commission
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
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17Work sponsored by theUS Nuclear Regulatory Commission
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
Irradiated Stainless Steels 289°C
Open Symbols: NWC BWR Env.Closed Symbols: HWC BWR Env.
Data from
Studsvik & Halden
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18Work sponsored by theUS Nuclear Regulatory Commission
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
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19Work sponsored by theUS Nuclear Regulatory Commission
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
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20Work sponsored by theUS Nuclear Regulatory Commission
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
Irradiated SSModel 8 ppm DO
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
Irradiated SSModel 8 ppm DO
Open Symbols: 0.75 dpaClosed Symbols: 2.16 dpa
SMA Weld HAZ
Sensitized
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21Work sponsored by theUS Nuclear Regulatory Commission
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
Irradiated SSModel 8 ppm DO
Specimen Number
Kmax
= 13 MPa m1/2
Irradiated SSModel 0.2 ppm DO
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22Work sponsored by theUS Nuclear Regulatory Commission
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
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23Work sponsored by theUS Nuclear Regulatory Commission
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
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24Work sponsored by theUS Nuclear Regulatory Commission
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)
Crack Extension (mm)
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
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25Work sponsored by theUS Nuclear Regulatory Commission
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)
Crack Extension (mm)
Type 304L SS, SAW HAZ
Fluence 1.44 x 1021 n/cm2
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)
Crack Extension (mm)
TYPE 304 SS, SMAW HAZ
Fluence 1.44 x 1021 n/cm2
289°C
Estimated Effective Flow Stress: 528 MPa
J = 219!a0.43
JIC = 128 kJ/m2
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26Work sponsored by theUS Nuclear Regulatory Commission
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)
Crack Extension (mm)
Type 304 SS (Heat 10285
Fluence 1.44 x 1021 n/cm2
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
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27Work sponsored by theUS Nuclear Regulatory Commission
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)
Crack Extension (mm)
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
Fluence 1.63 x 1021 n/cm2
289°C High-Purity Water
Specimen No.
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28Work sponsored by theUS Nuclear Regulatory Commission
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)
Neutron Exposure (dpa)
Austenitic SSs
ANL Heats
835 kJ/cm2
Closed Symbols: BWR WaterOpen Symbols: Air
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29Work sponsored by theUS Nuclear Regulatory Commission
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)
Neutron Exposure (dpa)
C = 20 + 205 exp(-0.65·dpa)
Cast Austenitic SSs & Welds
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30Work sponsored by theUS Nuclear Regulatory Commission
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)
Neutron Exposure (dpa)
C = 20 + 205 exp(-0.65·dpa)
Austenitic Stainless Steels
Data with X tested in BWR water
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31Work sponsored by theUS Nuclear Regulatory Commission
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)
Neutron Exposure (dpa)
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
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32Work sponsored by theUS Nuclear Regulatory Commission
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