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

1

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

1

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

1

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

Quad I

Quad IIQuad III

Quad IV

Median

Med

ian

1

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.