the effect of cold work on the crack response of dual...
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The effect of cold work on the crack response of dual certified 304/304L stainless steel containing boron
17th International Conference on Environmental Degradation of Materials in Nuclear Power Systems – Water Reactors, August 2015, Ottawa, Canada
Kevin Fisher1, Bryan Miller2, Earl Johns2, Bob Hermer2, Cathy Brown2, Emmanuelle Marquis1 1University of Michigan Department of Materials Science and Engineering 2Bechtel Marine Propulsion Corporation
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Background Results Conclusions Experimental Summary
17th International Conference on Environmental Degradation of Materials in Nuclear Power Systems – Water Reactors, August 2015, Ottawa, Canada
Typical SCC conditions
SCC
Material
Environment
Stress
• 304 SS • Sensitizing heat
treatments • Cr Carbide
Precipitates • GB Cr Depletion
• Cold Work • Residual
stress from welds
• High temperature water
• Aerated, deaerated • Impurities
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Background Results Conclusions Experimental Summary
17th International Conference on Environmental Degradation of Materials in Nuclear Power Systems – Water Reactors, August 2015, Ottawa, Canada
Towards SCC resistance
1. Decrease carbon content • 304L SS
– Regulations on other impurities
2.Reduce cold work level • How low is low enough?
– ~5% “threshold” for SCC? [1,2]
3.Eliminate sensitizing heat treatments • Not always feasible
4.Modify the environment • Remove oxygen, impurities?
[1] Raquet et al. 12th Environmental Degradation, 2005 [2] Arioka et al. Corrosion, 2007
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Background Results Conclusions Experimental Summary
17th International Conference on Environmental Degradation of Materials in Nuclear Power Systems – Water Reactors, August 2015, Ottawa, Canada
SCC under consideration
SCC
Material
Environment
Stress
• Dual Certified 304/304L SS, Mill annealed • Low B (<5
ppm) vs. High B (18 ppm)
• 0, 5, 10% CW
• 0.8” (C)T • Triangular load
500s rise/fall • Trapezoidal load
500s rise/fall, 9000 s hold
• 250°C • Aerated water + sulfates • Deaerated water
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Background Results Conclusions Experimental Summary
17th International Conference on Environmental Degradation of Materials in Nuclear Power Systems – Water Reactors, August 2015, Ottawa, Canada
Previous work on 304/304L with elevated B 1.Mill Anneal (1079ºC for 3.5 hours, water quenched)
• Fails ASTM A262 Practice A - oxalic acid etch test [3] • Passes ASTM A 262 Practice E – copper/copper sulfate/sulfuric acid
test [3] • TEM reveals GB precipitates [3]
– M2B – Cr content rarely dips below 12% in vicinity of precipitate
• DL-EPR indicates high degree of sensitization [4]
[3] Miller and Burke, 15th Environmental Degradation, 2011 [4] Johns and Miller, 16th Environmental Degradation, 2013
From [3]
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Background Results Conclusions Experimental Summary
17th International Conference on Environmental Degradation of Materials in Nuclear Power Systems – Water Reactors, August 2015, Ottawa, Canada
Experimental Approach
1.Cracking in 250°C water • Phase 1 – Triangular loading, aerated water + sulfates • Phase 2 – Trapezoidal loading, aerated water + sulfates • Phase 3 – Trapezoidal loading, deaerated water
2.Total, IG Crack Extension Measurements 3.EBSD Mapping
• Intergranular vs. Transgranular cracking • GB misorientation
4.APT and TEM • Oxide chemistry • Oxide/metal interface • GB chemistry
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Background Results Conclusions Experimental Summary
17th International Conference on Environmental Degradation of Materials in Nuclear Power Systems – Water Reactors, August 2015, Ottawa, Canada
Crack Extension and Exposure Low B High B
0 C
W
0 C
W
5 C
W
5 C
W
10 C
W
10 C
W
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Background Results Conclusions Experimental Summary
17th International Conference on Environmental Degradation of Materials in Nuclear Power Systems – Water Reactors, August 2015, Ottawa, Canada
EBSD
100 μm
100 μm
Low
B
Hig
h B
100 μm
100 μm
100 μm
100 μm
0% CW 5% CW 10% CW
38.5°
38.5° 30.3°
24.6°
47.7°
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Background Results Conclusions Experimental Summary
17th International Conference on Environmental Degradation of Materials in Nuclear Power Systems – Water Reactors, August 2015, Ottawa, Canada
Oxides using APT
No significant differences in chemistry were observed in different heats or regions
Stoichiometric Metal
Stoichiometric Oxygen
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Background Results Conclusions Experimental Summary
17th International Conference on Environmental Degradation of Materials in Nuclear Power Systems – Water Reactors, August 2015, Ottawa, Canada
Oxides using APT
No significant differences in chemistry were observed in different heats or regions
Stoichiometric Metal
Stoichiometric Oxygen
Stoichiometric Metal
Stoichiometric Oxygen
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Background Results Conclusions Experimental Summary
17th International Conference on Environmental Degradation of Materials in Nuclear Power Systems – Water Reactors, August 2015, Ottawa, Canada
Oxides using APT
No significant differences in chemistry were observed in different heats or regions
50 nm
Matrix Ni Cu Cr rich oxide
Stoichiometric Metal
Stoichiometric Oxygen
Stoichiometric Metal
Stoichiometric Oxygen
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Background Results Conclusions Experimental Summary
17th International Conference on Environmental Degradation of Materials in Nuclear Power Systems – Water Reactors, August 2015, Ottawa, Canada
Oxides using TEM/EDS – High B, 10%CW
GB Cr-rich oxide
3 um Fe-rich oxide
38.7
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Background Results Conclusions Experimental Summary
17th International Conference on Environmental Degradation of Materials in Nuclear Power Systems – Water Reactors, August 2015, Ottawa, Canada
Oxides using TEM/EDS – High B, 10%CW
GB Cr-rich oxide
FeCr2O4
3 um Fe-rich oxide
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Background Results Conclusions Experimental Summary
17th International Conference on Environmental Degradation of Materials in Nuclear Power Systems – Water Reactors, August 2015, Ottawa, Canada
Oxides using TEM/EDS – High B, 10%CW
GB Cr-rich oxide
FeCr2O4
Cr Ni
Fe O
3 um Fe-rich oxide
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Background Results Conclusions Experimental Summary
17th International Conference on Environmental Degradation of Materials in Nuclear Power Systems – Water Reactors, August 2015, Ottawa, Canada
Oxides using TEM/EDS – High B, 10%CW
GB Cr-rich oxide
FeCr2O4
Cr Ni
Fe O
3 um Fe-rich oxide
EDS results parallel APT results No significant Ni-rich region observed in GB ahead of crack Ni not consistently observed at oxide/metal interface No evidence of GB precipitates interacting with crack tip, despite proof they exist at
this grain boundary
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Background Results Conclusions Experimental Summary
17th International Conference on Environmental Degradation of Materials in Nuclear Power Systems – Water Reactors, August 2015, Ottawa, Canada
Oxides using TEM/EDS - High Boron, 10% CW
3 μm
Results are repeatable Despite several examples, yet to observe borides at the GB
Matrix Cr Rich Oxide Fe Rich Oxide
FeCr2O4
Cr-rich oxide
Fe-rich oxide
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Background Results Conclusions Experimental Summary
17th International Conference on Environmental Degradation of Materials in Nuclear Power Systems – Water Reactors, August 2015, Ottawa, Canada
What about GBs?– High B, 10%CW
‘Normal’ Grain Boundary
50 nm
Fe B
100 μm
38.5°
38.5°
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Background Results Conclusions Experimental Summary
17th International Conference on Environmental Degradation of Materials in Nuclear Power Systems – Water Reactors, August 2015, Ottawa, Canada
What about GBs?– High B, 10%CW
‘Normal’ Grain Boundary
50 nm
Fe B
100 μm
38.5°
38.5°
50 nm
Fe B
Grain Boundary Precipitation
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Background Results Conclusions Experimental Summary
17th International Conference on Environmental Degradation of Materials in Nuclear Power Systems – Water Reactors, August 2015, Ottawa, Canada
What about GBs?– High B, 10%CW
50 nm
Grain Boundary Near Precipitate
Fe B
‘Normal’ Grain Boundary
50 nm
Fe B
100 μm
38.5°
38.5°
50 nm
Fe B
Grain Boundary Precipitation
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Background Results Conclusions Experimental Summary
17th International Conference on Environmental Degradation of Materials in Nuclear Power Systems – Water Reactors, August 2015, Ottawa, Canada
GBs in Low B?– Low B, 10%CW
50 nm
Fe B
‘Normal’ Grain Boundary Compare to High B Grain Boundary
50 nm
Fe B
100 um
60.2°
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Background Results Conclusions Experimental Summary
17th International Conference on Environmental Degradation of Materials in Nuclear Power Systems – Water Reactors, August 2015, Ottawa, Canada
Summary of results 1. Oxide chemistry shows little variation across testing phase, type of crack, or
CW level • Cr-rich spinel adhered to crack flanks • Fe-rich magnetite particles filling open cracks
2. Some Ni and Cu enrichment observed at oxide/metal interfaces • Non-continuous at interface
3. GB chemistry in both high and low B materials were similar far away from any precipitates
• No precipitates observed in low B material • Slight enrichment of Cr at GB
4. GB along which cracks grew in high B material were decorated with chromium rich borides
• Cr depletion adjacent to borides was observed
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Background Results Conclusions Experimental Summary
17th International Conference on Environmental Degradation of Materials in Nuclear Power Systems – Water Reactors, August 2015, Ottawa, Canada
Conclusions 1. Boron in 304L can act similar to C, resulting in GB precipitation and Cr
depletion 2. GB boride precipitation alone is not sufficient to increase material susceptibly
to IGSCC • Synergistic effect with CW, starting around 5%, and environment/stress state • 0% CW specimen showed little to no evidence of intergranular SCC under same
conditions as 5% and 10% CW
Future Work 1. Determine the explicit role of boron in the cracking process
• What is happening to the borides as the GB oxidizes? 2. Differences in morphology of crack tip 3. What effect does sensitizing heat treatment have on the crack response and
precipitation propensity in high and low boron heats of dual certified 304/304L SS?
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Background Results Conclusions Experimental Summary
17th International Conference on Environmental Degradation of Materials in Nuclear Power Systems – Water Reactors, August 2015, Ottawa, Canada
Acknowledgements
This research was performed under appointment to the Rickover Fellowship Program in Nuclear Engineering sponsored by Naval Reactors Division of the U.S. Department of Energy.
Thank you! Questions?
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Additional Material
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Background Results Conclusions Experimental Summary
17th International Conference on Environmental Degradation of Materials in Nuclear Power Systems – Water Reactors, August 2015, Ottawa, Canada
EBSD – 0% CW Low Boron High Boron
100 μm 100 μm
Largely TG cracking driven by load form
100 μm 100 μm
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Background Results Conclusions Experimental Summary
17th International Conference on Environmental Degradation of Materials in Nuclear Power Systems – Water Reactors, August 2015, Ottawa, Canada
EBSD – 5% CW Low Boron High Boron
100 μm 100 μm
100 μm
Mixed IG and TG character, IG more prevalent in high B heat
100 μm
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Background Results Conclusions Experimental Summary
17th International Conference on Environmental Degradation of Materials in Nuclear Power Systems – Water Reactors, August 2015, Ottawa, Canada
Low B has IG branches, high B is almost exclusively IG IG Cracks propagate along high angle GB
EBSD – 10% CW Low Boron High Boron
100 μm 100 μm
100 μm
38.5°
38.5° 30.3° 24.6°
47.7°
100 um
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Background Results Conclusions Experimental Summary
17th International Conference on Environmental Degradation of Materials in Nuclear Power Systems – Water Reactors, August 2015, Ottawa, Canada
Total Crack Extension
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Background Results Conclusions Experimental Summary
17th International Conference on Environmental Degradation of Materials in Nuclear Power Systems – Water Reactors, August 2015, Ottawa, Canada
Experimental Approach Atom probe tomography
1. Use FIB to perform site specific lift out from area of interest
5 μm
2. Use FIB to prepare needle shaped specimens with feature of interest in apex
1 μm
Oxide
Matrix
Pt
200 nm
3. Electric field induces elemental evaporation from the needle’s apex 4. Time of flight mass spectrometry yields a mass spectrum. Identification of peaks allows for reconstruction of the original volume
Fe2+ Ni2+
O21+
CrO2+ FeO2+ NiO2+ Co2+ Mn2+
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Background Results Conclusions Experimental Summary
17th International Conference on Environmental Degradation of Materials in Nuclear Power Systems – Water Reactors, August 2015, Ottawa, Canada
Plan View TEM 1 (High B, 10%CW)
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Background Results Conclusions Experimental Summary
17th International Conference on Environmental Degradation of Materials in Nuclear Power Systems – Water Reactors, August 2015, Ottawa, Canada
Plan View TEM 3 (High B, 10%CW)
GB
Cr Rich oxide
Fe Rich oxide
Crack branch behind main crack. No evidence of borides here
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Background Results Conclusions Experimental Summary
17th International Conference on Environmental Degradation of Materials in Nuclear Power Systems – Water Reactors, August 2015, Ottawa, Canada
Cross Sectional TEM 1 (High B, 10%CW)
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