Advanced Radiation resistant Materials (ARRM) Program Update€¦ · Advanced Radiation resistant...
Transcript of Advanced Radiation resistant Materials (ARRM) Program Update€¦ · Advanced Radiation resistant...
Advanced Radiation resistant Materials (ARRM)Program Update
Mi Wang1, Miao Song1, Gary S. Was1, Lizhen Tan2,
Lawrence Nelson3, Raj Pathania4
1 University of Michigan2 Oak Ridge National Laboratory
3 JLN Consulting4 Electric Power Research Institute
Project sponsored by EPRI (contracts 10002164 and
10002154) and DOE (contract 4000136101)
International Light Water Reactors Material Reliability Conference and Exhibition, Chicago, IL, USA, August 1 – 4, 2016.
2© 2016 Electric Power Research Institute, Inc. All rights reserved.
Objectives of ARRM Program
EPRI, the U.S. Department of Energy (DOE) and Bechtel Marine
Propulsion Corp. (BMPC) have initiated a global, collaborative
research effort to develop the next generation of materials for in-
core structural components and fasteners
The two primary research goals are:
– By 2022, to develop and test a degradation-resistant alloy that is within
current commercial alloy specifications
– By 2024, to develop and test a new advanced alloy with superior
degradation resistance
3© 2016 Electric Power Research Institute, Inc. All rights reserved.
Alloy Score vs. State of Knowledge for Low-Strength Component Alloys
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2
3
4
5
8 10 12 14 16 18
Stat
e o
f K
no
wle
dge
Score
Candidate Alloys
Zr-2.5Nb Alloy 690 Alloy 600 Alloy 625
Hastelloy C22 Type 310 (23Cr-21Ni) Type 304/.../348 Alloy 800/825
Type 309 Zr alloys HT9/422/HCM12A 9Cr-RAFM
D9/AIM-1 HT-UPS Ti alloys Low Cr (2.25Cr-1Mo)
Higher Cr: >14Cr 12-14Cr ODS 8-9Cr ODS Mo alloys
Nb alloys High Cr, High Al ODS SiC-SiC Ta alloys
Ultrafine grain Multilayers Metal glasses W-Re-V alloys
TiAl ...
Median
Med
ian
Quad IV
Quad III
Quad I
Quad II
4© 2016 Electric Power Research Institute, Inc. All rights reserved.
Alloy Score vs. State of Knowledge for High-Strength Component Alloys
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2
3
4
5
8 10 12 14 16 18
Stat
e o
f K
no
wle
dge
Score
Candidate Alloys
Alloy 625 HS Alloy 725 Alloy X-750
Alloy 718 12-14Cr ODS (HS) 8-9Cr ODS (HS)
Quad I
Quad IIQuad III
Quad IV
Median
Med
ian
5© 2016 Electric Power Research Institute, Inc. All rights reserved.
Candidate Alloys for ARRM Program
Low Strength Applications
– 316L (Control)
– Alloy 625
– Alloy 690
– Alloy 800
– Alloy 310
– Zr-2.5Nb
– Grade 92 (9 Cr ferritic-martensitic steel)*
– HT9 (12 Cr ferritic-martensitic steel)*
High Strength Applications
– Alloy X-750 (Control)
– Alloy 625 plus
– Alloy 625 Direct Age
– Alloy 725
– Alloy 718
– 14YWT (14Cr oxide-dispersion-strengthened alloy)*
– Alloy 439 (18Cr ferritic stainless steel)*
* Advanced Alloys
6© 2016 Electric Power Research Institute, Inc. All rights reserved.
ARRM Alloy Testing Roadmap
Testing Stages 2011 2012 2013 2014 2015 2016 2017 2018 2019 2020 2021 2022 2023
Phase 0
Literature review
Critical Issues Report
Testing Phase 1
Procure Alloys and Characterize MIcrostructure
Non-Irradiated Materials Testing
Proton Irradiation/Testing
Ion Irradiation/Testing
Phase 1 Final Report/Select Alloys for Phase 2
Testing Phase 2
Non-Irradiated Material Testing
Neutron Irradiations
Low Dose (5-10 dpa) Neutron Testing
High Dose (20-40 dpa) Neutron Testing
Phase 2 Final Report/Identify Resistant Alloys
EPRI/DOE/BMPC Funding Phase 1 Funding Phase 2
Objective
• Select candidate alloys based on the IASCC behavior in both PWRprimary water and BWR-NWC environments.
• Understand the IASCC behavior based on the irradiation-inducedmicrostructure changes.
Candidate alloys
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8 10 12 14 16 18
Stat
e o
f K
no
wle
dge
Score
Candidate Alloys
Alloy 625 HS Alloy 725
Alloy X-750 Alloy 718
Quad I
Quad IIQuad III
Quad IV
Median
Med
ian
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2
3
4
5
8 10 12 14 16 18
Stat
e o
f K
no
wle
dge
Score
Candidate Alloys
Zr-2.5Nb Alloy 690 Alloy 600 Alloy 625 Hastelloy C22 Type 310 (23Cr-21Ni) Type 304/.../348 Alloy 800/825 Type 309 Zr alloys HT9/422/HCM12A 9Cr-RAFM D9/AIM-1 HT-UPS Ti alloys Low Cr (2.25Cr-1Mo) Higher Cr: >14Cr 12-14Cr ODS 8-9Cr ODS Mo alloys Nb alloys High Cr, High Al ODS SiC-SiC Ta alloys Ultrafine grain Multilayers Metal glasses
Median
Med
ian
Quad IV
Quad III
Quad I
Quad II
High-strength AlloyLow-strength Alloy
Status of experiment
Materials
Proton Irradiation
(5 dpa)
Microstructure
Characterization
CERT Test in PW &
BWR Analysis of Results
Alloy 625
Alloy 625 plus
Alloy 725
Alloy 625 Direct Age
Alloy 690
T92
316L
Alloy 800
Alloy 310
Alloy X-750
Zr-2.5Nb
C22
Alloy 439
Nickel-base alloysThe chemical composition and heat treatment had a significant impact on the microstructure of as-received
materials. The as-received materials were generally homogenous in these alloys.
625 625P625DA 725
Similar composition, different HT. Similar HT, different chemical compositions.
Alloy Ni Cr Mo Fe Nb Ti Al Ta C Si Mn
625 61.0 22.4 8.77 3.51 3.57 0.23 0.17 0.02 0.04 0.08 0.06
625DA 62.4 20.8 8.39 3.88 3.58 0.31 0.33 0.03 0.036 0.06 0.08
625P 60.4 21.0 8.02 5.76 3.4 1.28 0.2 - 0.008 0.03 0.02
725 57.6 21.5 8.07 7.94 3.41 1.35 0.17 0.01 0.011 0.04 0.04
Precipitates observed in alloy 725 (as-received condition)
50nm
V1
V2 V3
• γ'' phase was observed
• no γ' phase was observed
Density= 2.9 ×1022/m3
Mechanical properties at room temperature
Alloy Elastic
Modulus
(GPa)
Yield
stress
(MPa)
UTS
(MPa)
Plastic Strain
at fracture
625 227 400 868.1 0.454
625Plus 198 825 1198.9 0.310
625DA 213 867 1194.4 0.310
725 206 986 1296.4 0.289
Microstructural changes after proton irradiation
Experimental - proton irradiation
• 2 MeV proton irradiation at Michigan Ion Beam Laboratory (MIBL)
• Special design stage contains tensile and TEM samples
• Dose rate = ~ 1.3 x 10-5 dpa/s
• Irradiation temperature = 360 ± 5°C
• Final dose: ~ 5 dpa
0.00E+00
3.00E-04
6.00E-04
9.00E-04
1.20E-03
0 5 10 15 20 25
Dam
age
Rate
(d
pa/i
on
/an
g)
Target Depth (micron)
TRIM Damage Profile -- 2.0 MeV Protons on Alloy 725
γ" phase in alloy 725 after 5 dpa at 360°C
V1
V2 V3
Density (after irr)
= 2.3 ×1022/m3
Density (before irr)
= 2.9 ×1022/m3
Relrod of Frank loops in alloy 725 after 5 dpa at 360°C
100nm
Before irradiation
After irradiation
After irradiation
Density=1.4×1022/m3
Ni2Cr formation in alloy 725 after 5 dpa at 360°C
Three different variants of Ni2Cr
phase observed in the 112 direction
Density=2.0 ×1022/m3
Microstructure features in irradiated nickel-base alloys
Alloy γ" precipitates Disl. loops New formed phase(Ni2Cr)
Before irradiation After irradiation d(nm) ρ L d ρ fv
d ρ fv d(nm) ρ fv
625 N/A N/A N/A N/A N/A N/A 14.9±5.6 4.7±2.0 2.19 6.0±2.2 43±19 4.9
625P 13±3.3 4.5±3.5 2.51 12.7±3.4 6.3±1.6 3.3 16.7±5.4 3.1±0.8 1.62 6.1±2.1 10.5±2.6 1.2
725 18±6.0 2.9±1.2 3.69 18.9±4.7 2.3±0.9 3.2 23±8.7 1.4±0.6 1.01 7.8±2.6 2.0±0.9 0.5
625DA 12±3.1 2.1±0.2 0.78 11.5±3.3 6.6±4.0 2.7 13±5.8 10.2±1.1 4.16 10.8±3.8 9.6±3.9 6.3
Units: d, nm; ρ, 1022/m3; fv, %.; L, 1015/m2.
RIS in irradiated 625Plus
RIS tendency and GB chemistry
All Ni-base alloys exhibited Cr depletion and Ni enrichment at the grain boundary.
RIS was most severe in alloy 725 and least in alloy 625 Plus.
Irradiation hardening
Radiation induced hardening
Materials Hardness
(Hv)
Hardness after
irradiation (Hv)
Delta
Hardness
625 346 ± 24.5 447 ± 20.4 101
625Plus 424 ± 22.5 482 ± 24.4 58
625DA 420 ± 16.0 577± 28.0 157
725 428 ± 17.8 526 ± 18.8 98
316L 201±18.1 386±22.4 185
Evaluation of IASCC behavior
Experimental – CERT test
• Constant Extension Rate Tensile (CERT) test in both BWR(NWC) and PWR primary water environments
• slow strain rate: ~ 1 x 10-7 s-1
• strain to ~ 4%
Parameter BWR (NWC) PW
Temperature (∘C) 288 320
Pressure (psi) 1500 2000
Inlet Conductivity (µS/cm) < 0.1 20-30
Outlet Conductivity (µS/cm) < 0.1 30-30
O2 Concentration (ppb) 2000 < 5
H2 Concentration (cc/kg) - 35
[B] (ppm) - 900-1100
[Li] (ppm) - 2-3
pH at 25∘C 7.0 6 – 6.7
Stress-strain curve for alloy 725 irradiated to 5 dpa
Cracking on the irradiated surface in PWR primary water
• Surface covered with discontinuous heterogeneous oxide outer layer oxide
• IASCC cracks were observed for all Ni-base alloys, together with IGSCC cracks in the unirradiated area
Alloy 625 (IR) – 4.05%
Alloy 625Plus (IR) – 4.25% Alloy 725 -2 (IR) - 4.25%
10 μm
StressAlloy 625DA (IR) – 3.9%
Cracking on the irradiated surface in BWR-NWCAlloy 725 (IR) - 3.7% Alloy 625Plus (IR) – 3.6%
Alloy 625 (IR) – 3.45% Alloy 625DA (IR) – 4% Stress
10 μm
• Surface covered with discontinuous heterogeneous oxide outer layer oxide
• IASCC cracks were observed for all Ni-base alloys, together with IGSCC cracks in the unirradiated area
• Dislocation channels were observed on the irradiated area
SS 304* (commercial purity, 5 dpa), data comes from Jiao et al. (2011).
* Z. Jiao, G.S. Was, “Impact of localized deformation on IASCC in austenitic stainless steels”, Journal of Nuclear Materials 408 (2011) p. 246-256.
Comparison of cracking behavior between alloys in BWR-NWC
Comparison of cracking behavior between alloys in PWR primary water
Comparison of cracking between BWR-NWC and PWR primary water
PW NWC
Stephenson et al.*: on neutron irradiated stainless steels tested in CERT mode:
• %IG was higher in NWC than in primary water
• a higher density of non-propagating surface cracks was found in primary water
* K.J. Stephenson, G.S. Was, “Crack initiation behavior of neutron irradiated model and commercial stainless steels
in high temperature water”, Journal of Nuclear Materials 444 (2014) p. 331-241.
IASCC behavior and microstructure changes
725
625
625P
625DA
CL/UA
(μm/mm2)
IASCC (NWC)
1800
25000
725
625P
625DA
CL/UA
(μm/mm2)
IASCC (PW)
900
5700
625P
725
625
625DA
∆H/H
(%)
Hardening
14
37
625P
625DA
%Cr
at GB
(wt%)
RIS (Cr)
13.4
8.56
725
625P
625
625DA
fv
(%)
New phase (Ni2Cr)
0.5
6.3
725
625P
625DA
∆fv
(%)
∆γ" precipitates*
- 0.49
1.92
Dislocation Loops
Length
(1015/m2) 625
625DA
725
625P
1.01
4.16
*: no γ" precipitates in alloy 625.
725
625
625
Summary
• Alloy 725 has the lowest IASCC susceptibility among all the nickel-base alloys tested in both BWR-NWC and PWR primary water environments.
• Alloy 625DA has the highest IASCC susceptibility in both BWR-NWC and PWR primary water environments.
• Low strength alloy 625 is much more susceptible to IASCC than is 304 SS in BWR-NWC.
• Except for alloy 625, IASCC susceptibility of nickel-base alloys in BWR-NWC is higher than in PWR primary water.
• Irradiation induced microstructure changes are greatest in alloy 625DA, which also exhibits the highest susceptibility to IASCC in both BWR-NWC and PWR primary water environments.