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)

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Thanks for your kind attention!