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Structural Materials for Advanced Nuclear Systems
2016.05.02
Man Wang
Current Status of Structural Materials: 2nd Topic
Outline
1. Introduction of Fission Energy
2. Evolution of Advanced Nuclear Systems
3. Requirements for Materials
4. Candidate Structural Materials
2
31. Nuclear Fission Energy
Fast Neutron 1- 20 MeV
Thermal Neutron 0.025 eV
slowing by moderator
Sustainable fission chain reaction
4Nuclear Fission Reactor
fuel coolant moderator control rod
ceramic;metallic;
dispersion;liquid;
water;sodium;
gas;liquid metal;
water;graphite;
Boron;Ag-In-Cd;
1. Introduction of Fission Energy
2. Evolution of Advanced Nuclear Systems3. Requirements for Materials
4. Candidate Structural Materials
5
62. Evolution of Nuclear Fission Power
Generation Ⅳ International Form, 2002Improvement of Efficiency & Economics & Safety
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
1. Introduction of Fission Energy
2. Evolution of Advanced Nuclear Systems
3. Requirements for Materials4. Candidate Structural Materials
8
93. Serving Condition
High temp. & Radiation & Stress
①②
③
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!
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.
1. Introduction of Fission Energy
2. Evolution of Advanced Nuclear Systems
3. Requirements for Materials
4. Candidate Structural Materials
12
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
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
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
164.3 Ni-based Alloy
Advantages Traditional application at high temperature Good creep resistance
Disadvantages Irradiation brittlement Void swelling Phase instability due to irradiation
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.
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
19Fabrication
Pure metal element Powders
Yttrium Oxide
Yttrium Oxide
Pre-alloyed Gas Atomized Powders
OR
Y-Ti-O Y2O3
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
21Investigation of DispersoidsMA 956: Y-Al-O
MA 957: Y-Ti-O
22Mechanical Properties of ODS Alloy
Tensile test Creep Properties
23Irradiation Resistance of ODS Alloy
round cavities with small size
316-ODS
PNC 316
large faceted cavities
24Irradiation Resistance of ODS Alloy
Irradiation resistance can be improved by ODS!
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.
26
Thanks for your kind attention!