The Role of IASCC Material Degradation Processes on the ...
GE Hitachi Nuclear Energy / GE Global Research Center
The Role of IASCC Material Degradation Processes on the Assessment of BWR Core Shroud Long Term Operation P. L. Andresen and R. M. Horn International Light Water Reactor Materials Reliability Conference Chicago, IL
August 1-4, 2016
Outline Background on IASCC in BWRs Key Factors Affecting IASCC Management Evaluations of Core Shroud Cracking Summary
International LWR Material Reliability Conference Chicago, Il August 1-4, 2016
2
Background on IASCC in BWRs Irradiation starts to affect IGSCC crack growth rates of SS at ~1E20 n/cm2 – BWR industry accepts 5E20 n/cm2 for quantifiable changes BWRVIP efforts led to definition of IASCC growth rates for the range of 5E20 to 3E21 n/cm2 as discussed by Pathania, Carter, Horn and Andresen (2009) Integrated modeling is still needed to assess the crack length changes in regions of changing fluence, flux and chemistry GE GRC PLEDGE model is useful for these analyses It includes the effects of radiation segregation, hardening, residual stress relaxation, radiolysis/ECP, water purity, weld residual stresses, etc.
International LWR Material Reliability Conference Chicago, Il August 1-4, 2016
3
Standard Method to Assess IASCC in Core Shrouds Residual Stresses from welding
Core Shroud
Potential H4 2 in (51 m m ) type Residual Stress Distributions
Shroud Head Flange H1
60
Top Guide
50
W eld
H2
Top Guide Support Ring
40
Shroud
H3
30
H5 (present in som e plants)
Stress (ksi)
20
H4
Vessel W all
10
0
H6a
0
Core Plate
0.2
0.4
0.6
0.8
1
1.2
1.4
1.6
2
VIP14-4 50 ksi & 2 in
W eld
Core Plate Support Ring
1.8
NUREG-50 ksi & 2 in
-10
-20
H6b
VIP14-4 45 ksi & 2 in
-30
Shroud Support Plate
H7
VIP14-4 40 ksi & 2 in
-40 D epth (in)
Crack Growth Prediction
Material Behavior
(assum ing no sensitization and no cold w ork effects) 25
20 KW W Conductivity with 50 ksi Peak RS
Depth (mm)
15
KW W Case: EPR=0 uS/cm Neglible Cold work effects (M=1) Conductivity from plant ECP= 200 m V,she H4 Residual Stresses 50 ksi peak 0.5 ksi m em brane stress Flux = 2.5E19 n/cm 2/year
10
5 Unverified Results based on non Level 2 Code
0 0
50
100
150
200
250
300
350
400
Tim e (m onths)
Relies on disposition CGR and initial residual stresses International LWR Material Reliability Conference Chicago, Il August 1-4, 2016
4
Key Factors Affecting IASCC Management Modeling needs to account for the key factors affecting crack deepening Fluence level and flux in the region of cracking Local stress state from welding and dimensioning Changes in the grain boundary sensitization Water chemistry effects
All factors are changing with operational time International LWR Material Reliability Conference Chicago, Il August 1-4, 2016
5
Radiation Effects in Red
SCC Primary Variables • Corrosion Potential (esp. oxidants) • Water Purity – esp. Cl & SO4 • Yield Strength / Cold Work / Irradiation in bulk, surface or weld heat affected zone • Stress Intensity Factor – & cycling, vibration, dK/da • Sensitization (grain boundary Cr depletion) • Grain Boundary Carbides; Low Energy Boundaries • Temperature • Composition (Mo, Ti, Nb, low C, high N) not that important apart from decreasing sensitization International LWR Material Reliability Conference Chicago, Il August 1-4, 2016
6
Effect of Corrosion Potential
1.E-07
14.2 in/h
0.5 2000 ppb O2 Ann. 304SS 200 ppb O2
0.25
1.E-07
CW A600 42.5 28.3
14.2 in/h
CW A600
GE PLEDGE Predictions 30 MPam Sens SS
0.5
2000 ppb O2 Ann. 304SS 200 ppb O2
0.25 1.E-08
1.E-08
Includes: * All alloys tested * All heat treatments * No cold work data * Medium screening
Expected 16X peak in CGR for Alloy 182 vs. H2 in pure water
CGR data in pure water in Ar or low H2
0.06 S/cm 0.06 S/cm
-0.5
-0.4
-0.3
-0.2
-0.1
0.0
14.2 in/h
1.E-07
0.1
0.06 S/cm Industry Mean 30 MPam
42.5 28.3
1.E-08
0.1
1.E-09 -0.6
42000
1.E-06
316L (A14128, square ) 304L (Grand Gulf, circle ) non-sensitized SS 50%RA 140 C (black ) 10%RA 140C (grey )
Crack Growth Rate, mm/s
28.3
200
Circles = constant load Triangles = "gentle" cyclic SO4 & Cl data in pink
2000
Sens 304 SS Round Robin (open)
1.E-06
42.5
ppb O2
Alloy 182, Alloy 600 & St.Steel 30 MPam, 288C Water
2000 ppb O2
2000 ppb O2
Screened Round Robin data - highest quality data - corrected corr. potential - growth rates corrected to 30 MPam
GE PLEDGE Predictions 30 MPam
1.E-05
Sensitized 304 Stainless Steel 30 MPam, 288C Water 0.06-0.4 S/cm, 0-25 ppb SO4 SKI Round Robin Data filled triangle = constant load open squares = "gentle" cyclic
Crack Growth Rate, mm/s
Crack Growth Rate, mm/s
1.E-06
500 ppb O2
filled triangle = constant load open squares = "gentle" cyclic
200 ppb O2
1.E-05
Sensitized 304 Stainless Steel 30 MPam, 288C Water 0.06-0.4 S/cm, 0-25 ppb SO4
200 ppb O2 500 ppb O2
1.E-05
0.25 0.1 0.06 S/cm
SS PLEDGE Predictions * Normal YS
GE PLEDGE Predictions for Unsensitized Stainless Steel (upper curve for 20% CW) 0.1
Corrosion Potential, Vshe
0.2
0.3
0.4
1.E-09 -0.6
-0.5
-0.4
-0.3
-0.2
-0.1
0.0
0.1
Corrosion Potential, Vshe
0.2
0.3
0.4
1.E-09 -0.6
-0.5
-0.4
-0.3
-0.2
-0.1
0.0
0.1
0.2
0.3
0.4
Corrosion Potential, Vshe
SCC growth rate vs. corrosion potential for stainless steels (left and middle) and nickel alloys (right) tested in 288 C high purity water containing 2000 ppb O2 and 95 – 3000 ppb H2. International LWR Material Reliability Conference Chicago, Il August 1-4, 2016
7
200
150
0.4935
0.4930 100
Dissolved O2
0.4920
CT2 #7 - 304SS 4 dpa Constant Load, 19 ksiin Outlet Cond: 0.30 S/cm Inlet Cond: 0.27 S/cm Na2SO4
0.4915 1488
50
0 1508
1528
1548
1568
1588
Crack Growth Rate, mm/s
1.E-06
-6
Dissolved Oxygen, ppb
Crack Length, in
0.4940
0.4925
4 dpa 304SS
filled triangle = constant load open squares = "gentle" cyclic
-8
2.7 x 10 mm/s 1 x 10 mm/s
2000 ppb O2
0.4945
Sensitized 304 Stainless Steel 30 MPam, 288C Water 0.06-0.4 S/cm, 0-25 ppb SO4 SKI Round Robin Data
500 ppb O2
1.E-05 250
0.4950
200 ppb O2
IASCC Growth Rate Data
20% CW A600
316L (A14128, square ) 304L (Grand Gulf, circle ) non-sensitized SS 50%RA 140 C (black ) 10%RA 140C (grey ) 1.E-07
42.5 28.3
14.2 in/h
20% CW A600
GE PLEDGE Predictions 30 MPam Sens SS
0.5
2000 ppb O2 Ann. 304SS 200 ppb O2
0.25
0.1 S/cm Means from analysis of 120 lit. sens SS data 0.06 S/cm 0.06 S/cm
1.E-08
0.1
1608
Test Time, hr
At high potential, Cr depletion & radiation hardening affect SCC. At low potential, only hardening affects SCC. 1.E-09 -0.6
GE PLEDGE Predictions for Unsens. SS (upper curve for 20% CW) -0.5
-0.4
-0.3
-0.2
-0.1
0.0
0.1
0.2
0.3
0.4
Corrosion Potential, Vshe International LWR Material Reliability Conference Chicago, Il August 1-4, 2016
8
IASCC Radiation Creep Relaxation
Radiation creep relaxation of constant displacement stress (for uniform n-flux over the stressed area) produces large relaxation at relatively low fluence. If SCC doesn’t occur early, high fluence areas can be low in susceptibility. International LWR Material Reliability Conference Chicago, Il August 1-4, 2016
9
IASCC Radiation Induced Segregation (RIS) Radiation induced segregation is less reproducible than hardening or relaxation because or initial GB enrichment, composition variations, etc. Its origin is the migration of vacancies and interstitials that are created by displacement damage. SCC responds to minimum in Cr concentration profile, even when the Cr profile width is ~5 nm.
International LWR Material Reliability Conference Chicago, Il August 1-4, 2016
10
Fluence Effects on Fracture Toughness
Fracture toughness shows a large top between ~1 and ~20 dpa then an apparent saturation, as least out to ~100 dpa International LWR Material Reliability Conference Chicago, Il August 1-4, 2016
11
Example of Cracking Locations & Profiles
Crack Depth vs Azimuth
Cracks vs Fluence
Need understanding of IASCC behavior in high & low flux regions International LWR Material Reliability Conference Chicago, Il August 1-4, 2016
12
Effect of Neutron Fluence on Crack Growth Rates for Initially Un-sensitized SS Crack Growth Rate, mm/s
1.E‐06
Calculations for : Stainless steel, 10 ksiin 175 mVshe , 0.1 uS/cm Initially unsensitized 1.E‐07
1.E‐08
ECP Estimate for ~3X Benefit of OLNC at Low Fluence 1.E‐09 1.E+20
1.E+21 Neutron Fluence, n/cm2 (E>1 MeV)
The effect of RIS and radiation hardening that occur as the material fluence level changes between 1E20 and 2E21 n/cm2 leads to ~100 fold increase in CGR even with an effective HWC environment. At higher fluence levels, the IASCC crack growth rates are high. International LWR Material Reliability Conference Chicago, Il August 1-4, 2016
13
Relative Effect of Neutron Fluence on IASCC Crack Growth Rates in NWC 0.14
Fluence, dpa
1.43 120
Crack Growth Rate, mm/s
1.E‐06
Calculations for : Stainless steel, 10 ksiin 175 mVshe , 0.1 uS/cm
100
80
1.E‐07
60
40
1.E‐08
% steady state growth rate at constant K 1.E‐09 1.E+20
20
0 1.E+21 Neutron Fluence, n/cm2 (E>1Mev)
The green curve shows the fraction of ‘high fluence’ growth rate (on a linear scale) that exists at lower fluences. The largest change in growth rate occurs between about 0.5 – 2.5 dpa (upper axis). International LWR Material Reliability Conference Chicago, Il August 1-4, 2016
14
Evaluations of Core Shroud Cracking Assessments of long term effects of crack growth in BWR components requires a dynamic modeling capability Requires the ability to integrate material, stress & environmental effects GE GRC PLEDGE model has these capabilities Assumes active IGSCC/IASCC Benchmarked with early laboratory and plant data Modeling provides the ability to extrapolate with different assumptions Material condition, fluence and flux, water chemistry, and local and global stress conditions
PLEDGE modeling used for parametric evaluations International LWR Material Reliability Conference Chicago, Il August 1-4, 2016
15
Core Shroud Cracking Assessment Fluence Azimuthal Variation around the Core Shroud 3E21 n/cm2
2E21 n/cm2 1E21 n/cm2
Fluence Variation over the Core Shroud Cylinder
Azimuthal fluence variation drives future behavior International LWR Material Reliability Conference Chicago, Il August 1-4, 2016
16
Example Assessment: Existing Cracks in High Fluence Regions SS Material in high fluence regions already are established susceptibility Residual stresses have also relaxed due to radiation effects Local imposed fit-up stresses can be assumed to be present Changes that could be considered are: Mitigated environment (OLNC+H2) Modification to the Core Fuel configuration reducing flux Local surface mitigation to prevent crack lengthening
PLEDGE provides model for making assessment International LWR Material Reliability Conference Chicago, Il August 1-4, 2016
17
Example Assessment 1: Reduced Flux via Core Change 2
5 ksi fixed fit‐up stress, NWC, 0.1 S/cm, unsensitized SS. Limited benefit of 3X reduction in neutron flux on a 0.1" deep crack at 2.4 x 1021 n/cm2
1.8 1.6
Neutron flux lowered by 3X to 2.5 x 1019 n/cm2/h at 384 months
Crack Depth, inch
1.4 1.2
Neutron flux constant at 7.5 x 1019 n/cm2/y throughout life
1 0.8
Start with 0.1" deep crack at 2.4 x 1021 n/cm2 fluence at 384 months which is grown under different subsequent neutron flux levels.
0.6 0.4 0.2 0 350
360
370
380
390
400
410
420
430
440
450
Time, months
Conclusion: Once high fluence has been achieved, the effects of radiation on SCC are close to saturated; little further difference in crack trajectory International LWR Material Reliability Conference Chicago, Il August 1-4, 2016
18
Example Assessment 2: Reduced Flux via Core Change 2
5 ksi fixed fit‐up stress, NWC, 0.1 S/cm, unsensitized SS. Limited benefit of 3X reduction in neutron flux on a 0.1" deep crack at 1 x 1021 n/cm2
1.8 1.6
Neutron flux lowered by 3X to 1.04 x 1019 n/cm2/h at 384 months
Crack Depth, inch
1.4 1.2
Neutron flux constant at 3.125 x 1019 n/cm2/y throughout life
1 0.8
Start with 0.1" deep crack at 1.0 x 1021 n/cm2 fluence at 384 months which is grown under different subsequent neutron flux levels.
0.6 0.4 0.2 0 300
350
400
450
500
550
600
650
700
Time, months
Conclusion: Intermediate fluence locations are not saturated, mitigation or core management can have a big effect International LWR Material Reliability Conference Chicago, Il August 1-4, 2016
19
Summary of Factors Affecting Long Term Operation The known effects of radiation are key to assessment of existing flaws Fluence (as well as the flux) are key to material susceptibility and stress magnitudes The behavior of vertical cracks in welds regions is different than for horizontal cracks Far-field fit-up stresses will not relax Material susceptibility and toughness will saturate at high fluence Efforts to reduce flux in high fluence regions will have much less impact than continued environmental mitigation on future crack depth growth Through wall cracking at vertical welds more likely, impacting leakage assessments Role of surface residual stress relaxation on crack lengthening an open question
International LWR Material Reliability Conference Chicago, Il August 1-4, 2016
20
Summary of Crack Growth Modeling Benefits With SCC and IASCC, there are many influential and interacting factors that can affect the crack growth “trajectory” These include the following: Crack orientation/configuration, depth/length growth Current fluence, future neutron flux Type and origin of stress: weld residual stresses vs. far-field fit-up stress Corrosion potential, water purity Parametric modeling allows assessment of different future scenarios There are future benefits of IASCC crack lengthening models as well
International LWR Material Reliability Conference Chicago, Il August 1-4, 2016
21
The Role of IASCC Material Degradation Processes on the Assessment of BWR Core Shroud Long Term Operation P. L. Andresen and R. M. Horn International Light Water Reactor Materials Reliability Conference Chicago, IL
August 1-4, 2016
Outline Background on IASCC in BWRs Key Factors Affecting IASCC Management Evaluations of Core Shroud Cracking Summary
International LWR Material Reliability Conference Chicago, Il August 1-4, 2016
2
Background on IASCC in BWRs Irradiation starts to affect IGSCC crack growth rates of SS at ~1E20 n/cm2 – BWR industry accepts 5E20 n/cm2 for quantifiable changes BWRVIP efforts led to definition of IASCC growth rates for the range of 5E20 to 3E21 n/cm2 as discussed by Pathania, Carter, Horn and Andresen (2009) Integrated modeling is still needed to assess the crack length changes in regions of changing fluence, flux and chemistry GE GRC PLEDGE model is useful for these analyses It includes the effects of radiation segregation, hardening, residual stress relaxation, radiolysis/ECP, water purity, weld residual stresses, etc.
International LWR Material Reliability Conference Chicago, Il August 1-4, 2016
3
Standard Method to Assess IASCC in Core Shrouds Residual Stresses from welding
Core Shroud
Potential H4 2 in (51 m m ) type Residual Stress Distributions
Shroud Head Flange H1
60
Top Guide
50
W eld
H2
Top Guide Support Ring
40
Shroud
H3
30
H5 (present in som e plants)
Stress (ksi)
20
H4
Vessel W all
10
0
H6a
0
Core Plate
0.2
0.4
0.6
0.8
1
1.2
1.4
1.6
2
VIP14-4 50 ksi & 2 in
W eld
Core Plate Support Ring
1.8
NUREG-50 ksi & 2 in
-10
-20
H6b
VIP14-4 45 ksi & 2 in
-30
Shroud Support Plate
H7
VIP14-4 40 ksi & 2 in
-40 D epth (in)
Crack Growth Prediction
Material Behavior
(assum ing no sensitization and no cold w ork effects) 25
20 KW W Conductivity with 50 ksi Peak RS
Depth (mm)
15
KW W Case: EPR=0 uS/cm Neglible Cold work effects (M=1) Conductivity from plant ECP= 200 m V,she H4 Residual Stresses 50 ksi peak 0.5 ksi m em brane stress Flux = 2.5E19 n/cm 2/year
10
5 Unverified Results based on non Level 2 Code
0 0
50
100
150
200
250
300
350
400
Tim e (m onths)
Relies on disposition CGR and initial residual stresses International LWR Material Reliability Conference Chicago, Il August 1-4, 2016
4
Key Factors Affecting IASCC Management Modeling needs to account for the key factors affecting crack deepening Fluence level and flux in the region of cracking Local stress state from welding and dimensioning Changes in the grain boundary sensitization Water chemistry effects
All factors are changing with operational time International LWR Material Reliability Conference Chicago, Il August 1-4, 2016
5
Radiation Effects in Red
SCC Primary Variables • Corrosion Potential (esp. oxidants) • Water Purity – esp. Cl & SO4 • Yield Strength / Cold Work / Irradiation in bulk, surface or weld heat affected zone • Stress Intensity Factor – & cycling, vibration, dK/da • Sensitization (grain boundary Cr depletion) • Grain Boundary Carbides; Low Energy Boundaries • Temperature • Composition (Mo, Ti, Nb, low C, high N) not that important apart from decreasing sensitization International LWR Material Reliability Conference Chicago, Il August 1-4, 2016
6
Effect of Corrosion Potential
1.E-07
14.2 in/h
0.5 2000 ppb O2 Ann. 304SS 200 ppb O2
0.25
1.E-07
CW A600 42.5 28.3
14.2 in/h
CW A600
GE PLEDGE Predictions 30 MPam Sens SS
0.5
2000 ppb O2 Ann. 304SS 200 ppb O2
0.25 1.E-08
1.E-08
Includes: * All alloys tested * All heat treatments * No cold work data * Medium screening
Expected 16X peak in CGR for Alloy 182 vs. H2 in pure water
CGR data in pure water in Ar or low H2
0.06 S/cm 0.06 S/cm
-0.5
-0.4
-0.3
-0.2
-0.1
0.0
14.2 in/h
1.E-07
0.1
0.06 S/cm Industry Mean 30 MPam
42.5 28.3
1.E-08
0.1
1.E-09 -0.6
42000
1.E-06
316L (A14128, square ) 304L (Grand Gulf, circle ) non-sensitized SS 50%RA 140 C (black ) 10%RA 140C (grey )
Crack Growth Rate, mm/s
28.3
200
Circles = constant load Triangles = "gentle" cyclic SO4 & Cl data in pink
2000
Sens 304 SS Round Robin (open)
1.E-06
42.5
ppb O2
Alloy 182, Alloy 600 & St.Steel 30 MPam, 288C Water
2000 ppb O2
2000 ppb O2
Screened Round Robin data - highest quality data - corrected corr. potential - growth rates corrected to 30 MPam
GE PLEDGE Predictions 30 MPam
1.E-05
Sensitized 304 Stainless Steel 30 MPam, 288C Water 0.06-0.4 S/cm, 0-25 ppb SO4 SKI Round Robin Data filled triangle = constant load open squares = "gentle" cyclic
Crack Growth Rate, mm/s
Crack Growth Rate, mm/s
1.E-06
500 ppb O2
filled triangle = constant load open squares = "gentle" cyclic
200 ppb O2
1.E-05
Sensitized 304 Stainless Steel 30 MPam, 288C Water 0.06-0.4 S/cm, 0-25 ppb SO4
200 ppb O2 500 ppb O2
1.E-05
0.25 0.1 0.06 S/cm
SS PLEDGE Predictions * Normal YS
GE PLEDGE Predictions for Unsensitized Stainless Steel (upper curve for 20% CW) 0.1
Corrosion Potential, Vshe
0.2
0.3
0.4
1.E-09 -0.6
-0.5
-0.4
-0.3
-0.2
-0.1
0.0
0.1
Corrosion Potential, Vshe
0.2
0.3
0.4
1.E-09 -0.6
-0.5
-0.4
-0.3
-0.2
-0.1
0.0
0.1
0.2
0.3
0.4
Corrosion Potential, Vshe
SCC growth rate vs. corrosion potential for stainless steels (left and middle) and nickel alloys (right) tested in 288 C high purity water containing 2000 ppb O2 and 95 – 3000 ppb H2. International LWR Material Reliability Conference Chicago, Il August 1-4, 2016
7
200
150
0.4935
0.4930 100
Dissolved O2
0.4920
CT2 #7 - 304SS 4 dpa Constant Load, 19 ksiin Outlet Cond: 0.30 S/cm Inlet Cond: 0.27 S/cm Na2SO4
0.4915 1488
50
0 1508
1528
1548
1568
1588
Crack Growth Rate, mm/s
1.E-06
-6
Dissolved Oxygen, ppb
Crack Length, in
0.4940
0.4925
4 dpa 304SS
filled triangle = constant load open squares = "gentle" cyclic
-8
2.7 x 10 mm/s 1 x 10 mm/s
2000 ppb O2
0.4945
Sensitized 304 Stainless Steel 30 MPam, 288C Water 0.06-0.4 S/cm, 0-25 ppb SO4 SKI Round Robin Data
500 ppb O2
1.E-05 250
0.4950
200 ppb O2
IASCC Growth Rate Data
20% CW A600
316L (A14128, square ) 304L (Grand Gulf, circle ) non-sensitized SS 50%RA 140 C (black ) 10%RA 140C (grey ) 1.E-07
42.5 28.3
14.2 in/h
20% CW A600
GE PLEDGE Predictions 30 MPam Sens SS
0.5
2000 ppb O2 Ann. 304SS 200 ppb O2
0.25
0.1 S/cm Means from analysis of 120 lit. sens SS data 0.06 S/cm 0.06 S/cm
1.E-08
0.1
1608
Test Time, hr
At high potential, Cr depletion & radiation hardening affect SCC. At low potential, only hardening affects SCC. 1.E-09 -0.6
GE PLEDGE Predictions for Unsens. SS (upper curve for 20% CW) -0.5
-0.4
-0.3
-0.2
-0.1
0.0
0.1
0.2
0.3
0.4
Corrosion Potential, Vshe International LWR Material Reliability Conference Chicago, Il August 1-4, 2016
8
IASCC Radiation Creep Relaxation
Radiation creep relaxation of constant displacement stress (for uniform n-flux over the stressed area) produces large relaxation at relatively low fluence. If SCC doesn’t occur early, high fluence areas can be low in susceptibility. International LWR Material Reliability Conference Chicago, Il August 1-4, 2016
9
IASCC Radiation Induced Segregation (RIS) Radiation induced segregation is less reproducible than hardening or relaxation because or initial GB enrichment, composition variations, etc. Its origin is the migration of vacancies and interstitials that are created by displacement damage. SCC responds to minimum in Cr concentration profile, even when the Cr profile width is ~5 nm.
International LWR Material Reliability Conference Chicago, Il August 1-4, 2016
10
Fluence Effects on Fracture Toughness
Fracture toughness shows a large top between ~1 and ~20 dpa then an apparent saturation, as least out to ~100 dpa International LWR Material Reliability Conference Chicago, Il August 1-4, 2016
11
Example of Cracking Locations & Profiles
Crack Depth vs Azimuth
Cracks vs Fluence
Need understanding of IASCC behavior in high & low flux regions International LWR Material Reliability Conference Chicago, Il August 1-4, 2016
12
Effect of Neutron Fluence on Crack Growth Rates for Initially Un-sensitized SS Crack Growth Rate, mm/s
1.E‐06
Calculations for : Stainless steel, 10 ksiin 175 mVshe , 0.1 uS/cm Initially unsensitized 1.E‐07
1.E‐08
ECP Estimate for ~3X Benefit of OLNC at Low Fluence 1.E‐09 1.E+20
1.E+21 Neutron Fluence, n/cm2 (E>1 MeV)
The effect of RIS and radiation hardening that occur as the material fluence level changes between 1E20 and 2E21 n/cm2 leads to ~100 fold increase in CGR even with an effective HWC environment. At higher fluence levels, the IASCC crack growth rates are high. International LWR Material Reliability Conference Chicago, Il August 1-4, 2016
13
Relative Effect of Neutron Fluence on IASCC Crack Growth Rates in NWC 0.14
Fluence, dpa
1.43 120
Crack Growth Rate, mm/s
1.E‐06
Calculations for : Stainless steel, 10 ksiin 175 mVshe , 0.1 uS/cm
100
80
1.E‐07
60
40
1.E‐08
% steady state growth rate at constant K 1.E‐09 1.E+20
20
0 1.E+21 Neutron Fluence, n/cm2 (E>1Mev)
The green curve shows the fraction of ‘high fluence’ growth rate (on a linear scale) that exists at lower fluences. The largest change in growth rate occurs between about 0.5 – 2.5 dpa (upper axis). International LWR Material Reliability Conference Chicago, Il August 1-4, 2016
14
Evaluations of Core Shroud Cracking Assessments of long term effects of crack growth in BWR components requires a dynamic modeling capability Requires the ability to integrate material, stress & environmental effects GE GRC PLEDGE model has these capabilities Assumes active IGSCC/IASCC Benchmarked with early laboratory and plant data Modeling provides the ability to extrapolate with different assumptions Material condition, fluence and flux, water chemistry, and local and global stress conditions
PLEDGE modeling used for parametric evaluations International LWR Material Reliability Conference Chicago, Il August 1-4, 2016
15
Core Shroud Cracking Assessment Fluence Azimuthal Variation around the Core Shroud 3E21 n/cm2
2E21 n/cm2 1E21 n/cm2
Fluence Variation over the Core Shroud Cylinder
Azimuthal fluence variation drives future behavior International LWR Material Reliability Conference Chicago, Il August 1-4, 2016
16
Example Assessment: Existing Cracks in High Fluence Regions SS Material in high fluence regions already are established susceptibility Residual stresses have also relaxed due to radiation effects Local imposed fit-up stresses can be assumed to be present Changes that could be considered are: Mitigated environment (OLNC+H2) Modification to the Core Fuel configuration reducing flux Local surface mitigation to prevent crack lengthening
PLEDGE provides model for making assessment International LWR Material Reliability Conference Chicago, Il August 1-4, 2016
17
Example Assessment 1: Reduced Flux via Core Change 2
5 ksi fixed fit‐up stress, NWC, 0.1 S/cm, unsensitized SS. Limited benefit of 3X reduction in neutron flux on a 0.1" deep crack at 2.4 x 1021 n/cm2
1.8 1.6
Neutron flux lowered by 3X to 2.5 x 1019 n/cm2/h at 384 months
Crack Depth, inch
1.4 1.2
Neutron flux constant at 7.5 x 1019 n/cm2/y throughout life
1 0.8
Start with 0.1" deep crack at 2.4 x 1021 n/cm2 fluence at 384 months which is grown under different subsequent neutron flux levels.
0.6 0.4 0.2 0 350
360
370
380
390
400
410
420
430
440
450
Time, months
Conclusion: Once high fluence has been achieved, the effects of radiation on SCC are close to saturated; little further difference in crack trajectory International LWR Material Reliability Conference Chicago, Il August 1-4, 2016
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Example Assessment 2: Reduced Flux via Core Change 2
5 ksi fixed fit‐up stress, NWC, 0.1 S/cm, unsensitized SS. Limited benefit of 3X reduction in neutron flux on a 0.1" deep crack at 1 x 1021 n/cm2
1.8 1.6
Neutron flux lowered by 3X to 1.04 x 1019 n/cm2/h at 384 months
Crack Depth, inch
1.4 1.2
Neutron flux constant at 3.125 x 1019 n/cm2/y throughout life
1 0.8
Start with 0.1" deep crack at 1.0 x 1021 n/cm2 fluence at 384 months which is grown under different subsequent neutron flux levels.
0.6 0.4 0.2 0 300
350
400
450
500
550
600
650
700
Time, months
Conclusion: Intermediate fluence locations are not saturated, mitigation or core management can have a big effect International LWR Material Reliability Conference Chicago, Il August 1-4, 2016
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Summary of Factors Affecting Long Term Operation The known effects of radiation are key to assessment of existing flaws Fluence (as well as the flux) are key to material susceptibility and stress magnitudes The behavior of vertical cracks in welds regions is different than for horizontal cracks Far-field fit-up stresses will not relax Material susceptibility and toughness will saturate at high fluence Efforts to reduce flux in high fluence regions will have much less impact than continued environmental mitigation on future crack depth growth Through wall cracking at vertical welds more likely, impacting leakage assessments Role of surface residual stress relaxation on crack lengthening an open question
International LWR Material Reliability Conference Chicago, Il August 1-4, 2016
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Summary of Crack Growth Modeling Benefits With SCC and IASCC, there are many influential and interacting factors that can affect the crack growth “trajectory” These include the following: Crack orientation/configuration, depth/length growth Current fluence, future neutron flux Type and origin of stress: weld residual stresses vs. far-field fit-up stress Corrosion potential, water purity Parametric modeling allows assessment of different future scenarios There are future benefits of IASCC crack lengthening models as well
International LWR Material Reliability Conference Chicago, Il August 1-4, 2016
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