Structural Materials for

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

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1. Nuclear Fission Energy

Fast Neutron 1- 20 MeV Thermal Neutron

0.025 eV

slowing by moderator

Sustainable fission chain reaction

Nuclear 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 Systems 3. Requirements for Materials

4. Candidate Structural Materials

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2. Evolution of Nuclear Fission Power

Generation Ⅳ International Form, 2002

Improvement of Efficiency & Economics & Safety

Six Candidate Reactors – Gen Ⅳ

type coolant T_in / T_out (℃) 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 - GFR helium,

supercritical CO₂ 450 / 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 Materials

4. Candidate Structural Materials

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3. Serving Condition

High temp. & Radiation & Stress

② ①

③

Material 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!

Requirements 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

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Candidate Materials for Gen Ⅳ

type Cladding Structural 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 alloy High-Si austenitics, ceramics, refractory alloy MSR Not applicable Ceramics, refractory

metals, graphite, Ni alloy High-Mo, Ni-based alloy VHTR SiC or ZrC coating,

graphite Graphite, SiC, ZrC Ni-base superalloys, F/M

GFR ceramic Refractory metals,

ceramics, ODS Ni-base superalloys, F/M

SCWR F/M, ODS, Nickel alloy F/M, low alloy steel

4.1 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

4.2 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

4.3 Ni-based Alloy

Advantages

 Traditional application at high temperature

 Good creep resistance

Disadvantages

 Irradiation brittlement

 Void swelling

 Phase instability due to irradiation

4.4 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.

4.5 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

Fabrication

Pure metal element Powders

Yttrium Oxide

Yttrium Oxide

Pre-alloyed Gas Atomized P owders

OR

 Y-Ti-O

 Y₂O₃

Classification 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 ODS 9Cr-ODS

ODS Eurofer 97 9Cr research;

Japan, China, Europe

Austenitic ODS

304-ODS

based on austenitic steel

research;

China, Korean 316-ODS

310-ODS

Investigation of Dispersoids

MA 956: Y-Al-O

MA 957: Y-Ti-O

Mechanical Properties of ODS Alloy

Tensile test Creep Properties

Irradiation Resistance of ODS Alloy

round cavities with small size 316-ODS

PNC 316

large faceted cavities

Irradiation Resistance of ODS Alloy

Irradiation resistance can be improved by ODS!

Reference

[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!