Development of Advanced 9Cr Ferritic- Martensitic … Motivation for advanced structural materials...
Transcript of Development of Advanced 9Cr Ferritic- Martensitic … Motivation for advanced structural materials...
International Conference on Fast Reactors and Related Fuel Cycles: Safe Technologies and Sustainable Scenarios (FR13)
Paris, France
March 4-7, 2013
Development of Advanced 9Cr Ferritic-
Martensitic Steels and Austenitic Stainless
Steels for Sodium-Cooled Fast Reactors
T.-L. Sham, L. Tan and Y. Yamamoto
Oak Ridge National Laboratory, USA
Outline
Motivation for advanced structural materials
U.S. effort on SFR materials downselection
Advanced 9Cr ferritic-martensitic (FM) steels
– Tensile properties
– Thermal aging resistance
– Creep resistance
Advanced austenitic stainless steels
– Weldability
– Tensile properties
– Thermal stability
– Creep resistance
Summary
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Motivation for
Advanced Structural Materials
Ferritic-martensitic steels and
austenitic stainless steels are
primary structural materials of
SFRs.
Choosing the right materials can impact key requirements for advanced fast reactor development
– Economy: reduce capital costs through reduced commodities and simplifications
– Flexibility: higher material performance allows greater options to designers
– Safety: higher material performance promotes larger safety margins and more stable performance over longer lives
200 mm Dia.
593.3 C
32 MPa
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U.S. Advanced Materials Development
for Sodium Fast Reactors
2008
Established Alloy
Development Priority List
2009-2012
Alloys Downselection
2013-2015
Intermediate Term
Testing to Confirm
Enhanced Properties
• Considered a large class of
structural materials for further
development
• Involved 5 U.S. national
laboratories and 5 U.S.
universities
• Considered experience from
Fusion, Gen IV, Space
Reactor, and development
activities in Fossil Energy
• Established alloy
development priority list: ─ Ferritic-Martensitic steels
• Grade 92 (NF616)
• Grade 92 with thermo-
mechanical treatment (TMT)
─ Austenitic stainless steels
• HT-UPS
• NF-709
• Established comprehensive
downselection metrics
• Considered tensile properties, creep,
creep-fatigue, toughness, weldability,
thermal aging, sodium compatibility,
mechanical and TMT processes
• Integrated R&D activities by DOE Labs Oak Ridge National Laboratory
Argonne National Laboratory
Idaho National Laboratory
• Materials considered include Optimized-Gr92, Ta/Ti/V-modified 9Cr, Gr92,
Gr91 (baseline material)
HT-UPS (Fe-14Cr-16Ni), Modified HT-UPS,
NF709 (Fe-22Cr-25Ni), 316H (baseline
material)
• Based on overall performance w/
comprehensive metrics (and accelerated
test data), Optimized-Gr92 with TMT and
NF709 were downselected for further
assessment
• Further optimize
mechanical and TMT
processes
• Procure larger heats
• Validate performance
gains
• Longer-term testing of
base metals and
weldments
• Irradiation campaign
planning
• Development of
roadmap for ASME
nuclear code cases
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This Presentation
Focus on some aspects of downselection results
– Creep properties
– Resistance to thermal aging
– Weldability of austenitic stainless steels
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Advanced 9Cr FM Steels:
Tensile Properties
Compared to commercial Gr92, the advanced 9Cr FM steels had about 30-
60% increase in yield strength with about 5-40% reduction in total
elongation.
– The reduced elongations are still greater than the minimum requirement of Gr92
according to the ASTM standard A335/A213.
Δσy = (σ – σG92)/σG92
Δεt = (ε – εG92)/εG92
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σG92 – yield strength of Gr92
εG92 – total elongation of Gr92
Room-temperature tensile tests indicate that the Ta/Ti/V-modified heats showed
aging-induced softening. However, the yield strength of the Ti-modified heat was
partly recovered with an increased total elongation after 1,000 h aging.
The optimized-Gr92 with TMT (Gr92-2b) showed aging-induced hardening with a
slight reduction in total elongation.
Advanced 9Cr FM Steels:
Thermal Aging Resistance at 600oC
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100h
1,000h
100h
1,000h
100h
1,000h
100h
1,000h
Δσy = (σ – σG92)/σG92
Δεt = (ε – εG92)/εG92
σG92 – yield strength of Gr92
εG92 – total elongation of Gr92
Microstructures of the Aged
Optimized-Gr92 with TMT
The aging-induced softening was
primarily due to dislocation recovery.
– The dislocation density reduction would
reduce the strength by ~47% as
suggested by
The significantly increased ultrafine
precipitates after the aging at 600oC
for 1,000 h compensated the aging-
induced softening in the optimized-
Gr92 with TMT.
– The increased ultrafine particles would
increase the strength by ~70% as
suggested by the Orowan stress.
s r = 0.5MGb r f( )1 2
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Advanced 9Cr FM Steels:
Creep Resistance at 600oC
Compared to Gr91, the advanced 9Cr FM steels had significant increases in
creep life, especially for the optimized-Gr92 with TMT and the Ti-modified
heats with up to about 700 times increase.
= L
/LG
r91
600oC
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Alloy A HT-UPS
(Many small cracks at HAZ) (No cracks observed)
*Gas Tungsten Arc Welded, 11V, 180A, 8 inch/min, under Ar cover gas, applied 10% cold-roll prior to the welding.
1in
ch
1in
ch
Advanced Austenitic Alloys:
Weldability
Successful improvement of weldability of HT-UPS;
– Through careful control of alloying elements (both Alloys A and B).
– No defects on the surface and cross-sectional views.
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• No issues on
4t bend
specimens
• No visible
defects in
welds
Advanced Austenitic Alloys:
Weldability (cont’)
1” thickness plate weldments for further property evaluation
– Used double-V gas tungsten arc welding with the same filler material.
– Passed the standard inspection (4t bend test, ASTM E190).
– 10% cold-work (CW) applied to Alloy B prior to welding.
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Double-V Welding 4t Bend Test
Alloy B
NF709
0
1
2
3
4
0.1 1 10 100 1000 10000
Aging time, h 0
YS, 10% CW
HT-UPS
Alloy A
Alloy B
NF709
0
1
2
3
4
0.1 1 10 100 1000 10000
YS
re
lati
ve
to
re
fere
nc
e n
on
-CW
HT
-U
PS
Aging time, h 0
YS, no CW
Alloy B NF709
HT-UPS
Advanced Austenitic Alloys:
Tensile Properties at 650oC
Evaluated high temperature tensile properties after aging at 650oC.
– Base alloys showed a good stability of the properties .
– 10% cold work (CW) improved the yield strength, YS (greater than a factor of 2X of HT-
UPS with no CW).
• Original HT-UPS and Alloy B showed less thermal stability than the others after >1,000h aging at
650oC.
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0
20
40
60
80
100
650oC/200MPa
(no CW only)
(>100x)
0
100
200
300
400
500
Cre
ep
lif
e r
ela
tive t
o 3
16H
, x
10%CWno CW
(1x)
(1.6x)
700oC/200MPa
Advanced Austenitic Alloys:
Creep-rupture Properties
Strong Cold Work (CW) effect on improving creep-rupture life.
– HT-UPS showed the best properties among the candidates.
– NF709 exhibited the second best lives with/without CW.
– Alloy B showed better properties only when CW was applied.
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No
rmali
zed
cre
ep
str
ain
Relative time (linear)
Alloy B, 700oC, 200MPa no CW
10%CW
10%CW + Cross-weld
Advanced Austenitic Alloys:
Creep-rupture Properties (cont’)
Welding did not degrade the CW effect on the creep-properties (Alloy B).
– Effect of microstructural changes after welding was negligible.
– Literature reported that NF709 also showed no degradation of creep-properties after
welding (only for no cold work, CW).
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Summary of Creep, Thermal Aging
and Weldability Aspects
The creep resistance of advanced 9Cr FM steels was greatly enhanced by
optimizing their compositions as well as by using TMT.
– Up to about 700 times increase in creep life, compared to Gr91, was achieved under
the accelerated test conditions at 600oC.
The increased density of ultrafine precipitates facilitated the increase in
strength and thermal aging resistance, leading to the improved creep
resistance.
Properties of four candidate austenitic alloys, HT-UPS, NF709, and two
modified HT-UPS alloy (designated Alloys A and B), have been evaluated
and compared with 316H.
– Alloys A and B showed successful improvement in weldability.
– Only a little difference in thermal stability of the alloys in solution annealed conditions.
10% cold work increased the yield strength of the alloys for more than 200% compared
to the HT-UPS without cold work.
– HT-UPS exhibited the best creep properties among the alloys with and without cold
work, and NF709 followed.
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Acknowledgments & Copyright
Notice
This work is sponsored by the U.S. Department of Energy, Office of Nuclear
Energy, Advanced Reactor Concepts (ARC) Program, under contract DE-
AC05-00OR22725 with UT-Battelle, LLC.
This PowerPoint presentation has been authored by UT-Battelle, LLC, under
Contract No. DE-AC05-00OR22725 with the U.S. Department of Energy. The
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