structural materials for advanced nuclear systems · 2018. 4. 6. · structural materials for...
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Structural Materials for Advanced Nuclear Systems
2016.05.02
Man Wang
Current Status of Structural Materials: 2nd Topic
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Outline
1. Introduction of Fission Energy
2. Evolution of Advanced Nuclear Systems
3. Requirements for Materials
4. Candidate Structural Materials
2
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31. Nuclear Fission Energy
Fast Neutron 1- 20 MeV
Thermal Neutron 0.025 eV
slowing by moderator
Sustainable fission chain reaction
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4Nuclear Fission Reactor
fuel coolant moderator control rod
ceramic;metallic;
dispersion;liquid;
water;sodium;
gas;liquid metal;
water;graphite;
Boron;Ag-In-Cd;
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1. Introduction of Fission Energy
2. Evolution of Advanced Nuclear Systems3. Requirements for Materials
4. Candidate Structural Materials
5
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62. Evolution of Nuclear Fission Power
Generation Ⅳ International Form, 2002Improvement of Efficiency & Economics & Safety
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7Six Candidate Reactors – Gen Ⅳ
type coolant Tin / Tout (℃) Max. does/dpa
Supercritical water cooled reactor – SCWR
supercritical water 290 / 600 ~30
Very high temperature reactor - VHTR
helium 600 / 1000 <20
Gas fast reactor - GFRhelium,
supercritical CO2450 / 850 80
Sodium fast reactor- SFR sodium 370 / 550 200
Lead fast reactor - LFR Pb, Pb-Bi 600 / 800 150
Molten salt reactor - MSR molten salt 700 / 1000 200
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1. Introduction of Fission Energy
2. Evolution of Advanced Nuclear Systems
3. Requirements for Materials4. Candidate Structural Materials
8
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93. Serving Condition
High temp. & Radiation & Stress
①②
③
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10Material Limiting Phenomenon for Gen Ⅳ
1. High-temp. high does system: SFR, LFR, MSR strength, creep and creep-fatigue behavior void swelling and phase instability due to high level does
2. Very high-temp. gas cooled system: VHTR, GFR coolant (He) containing impurity: CO, CO2, CH4, H2O corrosion & oxidation
3. Supercritical water cooled system: SCWR supercritical water – 374℃/ 22 MPa stress corrosion cracking (SCC) irradiation assisted stress corrosion cracking (IASCC)
Materials!
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11Requirements for Materials
Resistance of irradiation embrittlement and swelling Good high temp. strength and creep resistance Corrosion & Oxidation resistance Low susceptibility to SCC Compatibility with coolant at high temp.
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1. Introduction of Fission Energy
2. Evolution of Advanced Nuclear Systems
3. Requirements for Materials
4. Candidate Structural Materials
12
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13Candidate Materials for Gen Ⅳ
type CladdingStructural Materials
In-core Out-of-core
SFR F/M, F/M ODS F/M, 316 SS ferritics, austenitics
LFR High-Si F/M, ODS, ceramics, refractory alloyHigh-Si austenitics,
ceramics, refractory alloy
MSR Not applicableCeramics, refractory
metals, graphite, Ni alloyHigh-Mo, Ni-based alloy
VHTR SiC or ZrC coating,graphite Graphite, SiC, ZrC Ni-base superalloys, F/M
GFR ceramicRefractory metals,
ceramics, ODSNi-base superalloys, F/M
SCWR F/M, ODS, Nickel alloy F/M, low alloy steel
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144.1 Ferrite / Martensitic Steel (9-12Cr)
austenitization → quenching → tempering at 760℃ferrite + martensite (F/M)
Advantages Better corrosion & oxidation
resistance Excellent reduced-activation Good swelling resistance
Disadvantages Low strength at high temp. Irradiation embrittlement
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154.2 Austenitic Stainless Steels
304 SS; 316 SS;
Advantages Good creep resistance
at high temp. Reasonable oxidation &
corrosion resistance
Disadvantages Severe void swelling Low thermal conductivity
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164.3 Ni-based Alloy
Advantages Traditional application at high temperature Good creep resistance
Disadvantages Irradiation brittlement Void swelling Phase instability due to irradiation
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174.4 Refractory Alloy
Advantages Good strength at high temperature Swelling resistance up to high burn ups
Disadvantages Poor oxidation resistance Fabrication difficulty Embrittlement at low temperature
Nb, Mo, Ta, etc.
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184.5 Oxide Dispersion Strengthening Alloy nano-sized dispersoids with high number density
→ strong pinning on dislocation movement→ excellent high temp. strength and creep resistance
interface between dispersoids and matrix→ sinks for defects→ improvement of irradiation resistance
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19Fabrication
Pure metal element Powders
Yttrium Oxide
Yttrium Oxide
Pre-alloyed Gas Atomized Powders
OR
Y-Ti-O Y2O3
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20Classification of ODS alloys
type character remark
Ferritic ODS
MA956 22Cr Commercial; USA
MA957 14Cr Commercial; USA
PM2000 18Cr Commercial; Germany
14YWT 14Cr research
12YWT 12Cr research
F/M ODS9Cr-ODS
ODS Eurofer 979Cr
research;Japan, China, Europe
Austenitic ODS
304-ODSbased on austenitic
steelresearch;
China, Korean316-ODS
310-ODS
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21Investigation of DispersoidsMA 956: Y-Al-O
MA 957: Y-Ti-O
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22Mechanical Properties of ODS Alloy
Tensile test Creep Properties
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23Irradiation Resistance of ODS Alloy
round cavities with small size
316-ODS
PNC 316
large faceted cavities
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24Irradiation Resistance of ODS Alloy
Irradiation resistance can be improved by ODS!
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25Reference[1] T. Abram, S. Ion, Energy Policy 36(12) (2008) 4323-4330.[2] J. Li, W. Zheng, S. Penttilä, et al., J. Nucl. Mater. 454(1-3) (2014) 7-11.[3] S.J. Zinkle, G.S. Was, Acta Mater. 61(3) (2013) 735-758.[4] K.L. Murty, I. Charit, J. Nucl. Mater. 383(1-2) (2008) 189-195.[5] D.A. McClintock, D.T. Hoelzer, M.A. Sokolov, et al., J. Nucl. Mater. 386-
388 (2009) 307-311.[6] D.A. McClintock, M.A. Sokolov, D.T. Hoelzer, et al, J. Nucl. Mater. 392(2)
(2009) 353-359.[7] H. Oka, M. Watanabe, H. Kinoshita, et al., J. Nucl. Mater. 417(1-3)
(2011) 279-282.[8] S. Ukai, M. Fujiwara, J. Nucl. Mater. 307 (2002) 749-757.[9] S. Ukai, S. Mizuta, M. Fujiwara, et al., J. Nucl. Sci. Technol. 39(7) (2002)
778-788.
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Thanks for your kind attention!