Australian’s Molten Salt Reactor (MSR) Material Research Ondrej Muránsky

Australian’s Molten Salt Reactor (MSR) 

Material Research 

Ondrej Muránsky

Australia’s Resources

Uranium in Australia 

• Has the world’s largest 

reserves of uranium 

(23% of the world total) 

• Is the world's thirdranking 

producer behind 

Kazakhstan and Canada 

• In 2012, produced over 

7000 tonnes of uranium 

oxide concentrate 

(U 3 O 8 ) 

• Exports to countries who 

have signed the Nuclear 

Non-Proliferation Treaty. 

Source: Geoscience Australia

• Carbonatite, Placer 

depositions, Vin-type 

deposits, Alkaline rocks: 

World resources: ~ 6 mil 

tons. 

• Australian Thorium 

resources estimate ~0.5 

mil tons. 

• ~ 8.8% of total world 

resources 

• No current production of 

thorium in Australia 

Thorium in Australia 

Source: Geoscience Australia

ANSTO 

Australian Nuclear Science and 

Technology Organisation

Australian’s Nuclear Lab for 50 Years 

AAEC 

1952 - 1987 

ANSTO 

1987 - Today

New ANSTO Research Structure 

ANSTO Gen IV and MSR Research 

7

Australia and the Gen IV International Forum 

• In 2015, the Australia Federal 

Government petitioned to join the 

Generation IV International Forum 

• Petition presented to GIF Policy 

Group in Oct 2015 

• GIF Policy Group Delegation Visit to 

Sydney in Feb 2016 

• GIF Policy Group unanimously 

endorsed Australian membership in 

April, 2016 

• ANSTO signed the GIF Charter in 

June 2016 initiating Australia’s 

membership into the Forum 

GIF Policy Group Delegation to 

Australia, 2-4 February 2016 

Prof Lyndon Edwards, former 

Head, Institute of Materials Engineering now 

National Director, Australian Generation IV 

International Forum Research

Australian key capability overview 

• Reactor risk and safety analysis 

• Radiation damage 

• Creep, creep fatigue and code development 

• Molten salt corrosion and testing 

• Atomistic and molecular modelling 

• Structural Integrity and weld modelling 

Also some expertise on GIF relevant: 

• Education and Training 

• Economic Modelling (principally from UNSW)

Australia’s GIF Research 

Following its success in gaining membership of GIF, Australia is 

proposing to join: 

• The Very High Temp Reactor (VHTR) SA 

• The Molten Salt Reactor (MSR) pSSC 

• The VHTR Material (MAT) PA 

• The Risk and Safety Working Group. 

In addition, with our University partners we are exploring how we 

might make contributions to: 

• The GIF Education and Training Working Group 

• The Economic Modelling Working Group 

SA - System Arrangements 

pSSC - Provisional System Steering Committee 

PA - Project Arrangements

Molten Salt Reactor (MSR) 

Material Research

ANSTO: Nuclear Fuel Cycle R&D 

• Fuel/cladding interactions 

– Use of atomistic modelling (e.g. DFT) 

• Structural materials performance under extreme 

conditions 

– Irradiation, corrosion, high temperatures and/or deformation 

• Separation science 

– Synthesis and analysis of titania frameworks for separation of U, Pu 

• Wasteform fabrication 

– Construction of a Synroc waste treatment plant to reduce volume of ANSTO 

nuclear by-products by 99%

ANSTO/SINAP TMSR Research Program 

• Shanghai Institute of Applied Physics 

(SINAP), Chinese Academy of 

Science (CAS) 

– Centre for Thorium Molten Salt 

Reactor (TMSR) systems 

• Australia/China Science and 

Research Fund Grant 2013-2015 

• ANSTO-SINAP Joint Materials 

Research Centre 

– Materials technology for Molten Salt 

Reactors 

• Molten salt corrosion 


• High temperature behaviour

ANSTO/SINAP TMSR Research Program 

• Shanghai Institute of Applied Physics 

(SINAP), Chinese Academy of 

Science (CAS) 

– Centre for Thorium Molten Salt 

Reactor (TMSR) systems 

• Australia/China Science and 

Research Fund Grant 2013-2015 

• ANSTO-SINAP Joint Materials 

Research Centre 

– Materials technology for Molten Salt 

Reactors 

• Molten salt corrosion 


• High temperature behaviour

SINAP - ANSTO 

Material Development

SINAP Material Development 

- Property of Hastelloy N: 

- Very good corrosion resistance to molten salt 

- Operated successfully in MSRE more than four years 

- Still the best choice for the MSR structural material 

- Only 704℃ 

- Helium embrittlement 

Chao Yang, PhD 

Candidate, 

Prof. Xingtai Zhou 

12 months stay at 

ANSTO 

High-temperature 

strength 

Corrosion 

resistance 

Irradiation 

resistance 

Development of Ni-SiC and NiMo-SiC 

dispersion precipitation strengthened 

(DPS) alloys for future TMSR systems. 

Advantages of SiC: Ceramic material, 

Irradiation resistance, Corrosion 

resistance, High temperature stability 

F/M/A steel 

ODS steel 

Hastelloy N 

GH3535 

ODS alloy 

Ceramics 

- ODS nickel alloys(weak corrosion resistance ) 

- Ceramics & ceramic matrix composites(lack of connection 

technology )

SINAP Materials 

• Development of new materials for molten salt reactor 

systems environments (high-temperature, corrosion, 

radiation) 

SINAP Materials: 

1. Ni-SiC, Ni-16Mo-SiC 

2. GH3535, a Chinese variant of Hastelloy N with 

the nominal composition of Ni–16Mo–7Cr–4Fe 

and Si used as an O getter 

3. Various Grades of Nuclear Graphite.

SINAP NiMo-SiC Alloy 

SiC 

SiC 

Ni 3 Si 

• SiC – dispersion strengthening 

• Ni 3 Si – precipitation strengthening 

• Mo 2 C – grain refinement, degradation of elongation 

C. Yang, O. Muransky, H. Zhu, G.J. Thorogood, H. Huang, X. Zhou, On the origin of strengthening mechanics in Ni-Mo alloys 

prepared via powder metallurgy, Materials and Design, DOI:10.1016/j.matdes.2016.10.024.





Intensity [au] 

Lattice Spacing [A] 



Material Testing in Extreme 

Environments

ANSTO Material Testing 

• Irradiation 

– Neutron, ion damage studies 

• High temperature 

– Creep testing 

• Corrosion 

– Testing in molten salt environments 

• Combined environments 

– Irradiation + corrosion (MSR) 

– Corrosion + high temperature creep






• Corrosion 




– Corrosion + high temperature creep

Neutron Irradiation Specimens 

• Miniature dog-bone samples 

• Small punch discs 

• Compact tension discs 

• Quarter-size Charpy samples 

• Corrosion samples 

1 dpa/yr

Samples Currently in OPAL (or Planned for Insertion) 

Material Sample type Irradiation time Source 

MA957 TEM discs 460 days IME/ORNL i 

CLAM “ “ IME i 

TiAl FT “ “ IME/LANL ii 

AlMg5 Mini tensile “ IME/OPAL Surveillance iv 

Zr-3.5Sn “ 460 days IME/Queens ii 

TiAl DP TEM discs 460 days IME/LANL ii 

Ti-6Al-4V “ “ IME i 

Zr-2.5Nb “ “ IME i 

Zr-4 CT 460 days OPAL Surveillance iv 

Zr-4 CT 5 years OPAL Surveillance iv 


NiCrMoFe Small Punch 460 days IME/SINAP JRC iii 

NiCrMoFe “ 920 days IME/SINAP JRC iii 

Ti45Al “ 460 days IME i 

Ti45Al2Nb “ “ IME i 

Ti45Al TEM disc “ IME i 


Zr-2.5Nb “ “ IME/OPAL i 

NiCrMoFe “ “ IME/SINAP JRC iii 

Zr-3.5Sn Mini Tensile 5 years IME/Queens ii 


Al6061 “ 5 years OPAL Surveillance iv 

i 

Internal research 

ii 

unfunded collaborative research 

iii 

partially funded research with external partners 

iv 

OPAL Materials Surveillance Program (MSP) 

• Zr and Al alloys 

– Predominantly for OPAL support 

• Titanium Aluminides 

– Internal research 

• Ni alloys 

– Collaborative work (SINAP MSR) 

• Additional planned materials 

– A508 (PM, WM, HIP) 

– ODS 

– Graphite 

– Tungsten 

26

Active Sample Machining and Testing 

• Active samples sent to MEHC, which contains: 

– Mechanical testing facilities for active material 

– Micromachining facilities to allow samples to exit cells 

• MEHC currently in licensing/commissioning stage 

27

Ion Irradiation at ANSTO 

• 1MV VEGA accelerator 

• 2MV STAR tandetron accelerator 

• 6MV SIRIUS tandem accelerator 

• 10MV ANTARES tandem 

accelerator 

– Energies from < 1 MeV to 100 

MeV

Single Crystal Ni, He+ Ion Irradiation 

11.85 µm 

~ 19 dpa 

Irradiation 

direction 

Front surface 

(Entry surface) 

Back surface 

(Exit surface) 

~ 10 dpa 

SRIM (The Stopping and Range of Ions in Matter) estimates for He+ irradiation of Ni 

29

In-site 1MeV Kr + ion irradiation at RT & 450C 

Argonne National 

Laboratory Hitachi 

H9000NAR IVEM 

operated at 200 kV 

In-situ irradiation experiments were performed on TEM 

thinned (~ 100 nm) Ni-Mo-Cr-Fe alloys using 1 MeV Kr + 

ions at room temperature (25ᵒC) and 450ᵒC up to to 100 

dpa (1.77 x 10 18 ions/cm 2 ). [110] zone axis. 

M. Reyes et al., Microstructural Evolution of an Ion Irradiated Ni-Mo-Cr-Fe Alloy at Elevated Temperatures. 

Materials Transactions, 2014, 55, pp. 428-433.

Dislocation Structure 

• In-situ 1 MeV Kr + ion irradiation at 

450 o C 

• Post Facto TEM analysis 

• High densities of both faulted and unfaulted 

loops observed 

• Analysis shows they are un-faulted 

½〈110〉 SIA loops and faulted 1/3〈111〉 

vacancy Frank loops 

• MD simulations of defects near ½〈110〉 

edge dislocations suggests that 

redistribution of displaced atoms from 

the cascade region towards the 

dislocation core may result in 

dislocation climb so remaining 

vacancies may partially collapse and 

form 1/3〈111〉 faulted vacancy Frank 

loops nearby. 

b 

• Faulted Loop 

• Faulted Loop 

a 

• Un-Faulted Loop

In situ micro tensile testing of He +2 ion irradiated and implanted single crystal nickel film. 

Acta Materialia, 2015, 100, pp. 147-154. 

Micromechanical Tensile Testing 

Radiation direction

Trends in Radiation Strengthening with Dose 

• Ion irradiation provides useful qualitative information 

• Full-sample performance does not explicitly consider localised damage 

In situ micro tensile testing of He +2 ion irradiated and implanted single crystal nickel film. 

Acta Materialia, 2015, 100, pp. 147-154.






• Corrosion 




– Corrosion + high temperature 

– Irradiation under high temperature

SINAP GH3535 vs Hastelloy N 

Larson Miller Parameter (LMP) 

LMP = T [log (t r ) + C] 

T - temperature ( o K) 

t r - time to rupture (hrs) 

C - constant (17.5)

SINAP GH3535 vs Hastelloy N 

Dorn-Shephard Parameter for 1% creep 

Log(1/ t 1%creep ) + Log(exp(42600/T) 

T - temperature ( o K) 

t 1%creep - time to 1% creep strain (hrs) 

Creep exponent 

n = 4






• Corrosion 






ANSTO Corrosion Testing 

• ANSTO-SINAP JRC (MSR) 

– FLiNaK salt composition 

– Tests conducted at 650ºC - 750ºC for 

10, 100 and 200 h 

• SINAP Materials 

– GR 3535, Graphite, NiMo-SiC, 

Hastelloy-N 

Hastelloy-N, Electron Back-Scatter Diffraction (EBSD) and 

Energy Dispersive Spectroscopy (EDS) chromium (Cr) 

distribution map, 200h/650C, FLiNaK.

Corrosion/Salt Infiltration into Graphite 

• Salt infiltration into various grades of graphite at 

different pressures (a) 1.0 atm; (b) 1.5 atm; (c) 3.0 atm; 

and (d) 5.0 atm. 

Z. He et al., Molten FLiNaK salt infiltration into degassed nuclear 

graphite under inert gas pressure, Carbon, 2015, 84, pp. 511-518.






• Corrosion 




– Corrosion + high temperature

SINAP GH3535: Irradiation & Corrosion 

GH3535: Vapour etch top figures, salt corrosion bottom figures 

H. Zhu et al. High-temperature corrosion of helium ion-irradiated Ni-based alloy in fluoride molten salt 

Corrosion Science, 2015, 91, pp. 1-6.

SINAP GH3535: Irradiation & Corrosion 

As-Received 

Irradiated, No Corrosion 

Corrosion, No Radiation 

Irradiated + Corrosion 

• GH3535 alloy, FLiNaK salt, 10 17 ions/cm 2 He+ 

• Helium ion irradiation increases the thickness of the corrosion layer in the 

irradiated and corroded sample to more 30 times than in the un-irradiated sample. 

H. Zhu et al. High-temperature corrosion of helium ion-irradiated Ni-based alloy in fluoride molten salt 

Corrosion Science, 2015, 91, pp. 1-6.

SINAP Graphite: Irradiation & Corrosion 

Graphite, FLiNaK salt, 10 17 ions/cm 2 He+ 

As-Received 




• Un-irradiated 

• Irradiated 

• Graphite test coupons were irradiated with 30 keV He+ ions 

• One section was masked to prevent radiation damage 

• The samples exposed to molten FLiNaK salt (150h@750 ºC). 

• Clear variations in the corroded structures are visible 

• NEXAFS suggests Fluorination of surface 

Damaged 

Masked






• Corrosion 




– Corrosion + high temperature creep 


SINAP GH 3535 Creep Testing in Molten Salt 

GH3553 creep rupture test (190 MPa @ 700 ºC) 

Initial results show significant effect of salt on creep life


SEM Band contrast Cr Mo 

20 µm 

EDS chemical 

composition 

(Average wt%) 

Region Ni Mo Cr Fe Mn Si 

Bulk sample 71.1 16.3 7.3 4.1 0.7 0.5 

Salt affected regions (st dev) 82.8 (0.5) 12.1 (0.4) 1.5 (0.2) 3.2 (0.2) 0.2 (0) 0.1 (0)


SEM 

Euler Map 

Cr 

20 µm 

Evidence of preferential 

grain boundary attack

SINAP GH 3535 Weld Characterisation 

48

ANSTO’s Numerical Weld 

Modelling

Prediction of Weld Residual Stress 

- ANSTO has developed complex 

models of microstructure and 

stress state around multi-pass 

welded joints 

- Used to support plant 

maintenance and design 

decisions 

NeT TG6 three-pass Inconel 82 slot-weld in an Inconel 

600 (international benchmark specimen). The Abaqus 

half model depicting the basic plate geometry and three 

consecutive passes filling the slot. The insert shows in 

detail elements associated with passes 1 to 3. 

76 mm (Z) 

200 mm (Z) 

150 mm (X)

Thermal Analysis 

PASS.3 

PASS.2 

PASS.1 

Welding Direction 

1380C Isotherm 

Weld Metal 

Parent Metal

Material Properties 

Comparison of the room temperature cyclic responses of Inconel 600 (parent metal) and 

Inconel 82 (weld metal) predicted by mixed isotropic-kinematic hardening plasticity 

model when using the first and second cycle to derive mixed hardening parameters.


Z 

X 

- Weld Metal: Inconel 82 

- Parent Metal: Inconel 600 

- 3-Pass Slot weld 

D Plane


D2 Line 

D10 Line

Thank you

High Temperature Irradiation (Planned) 

• Motivation 

– Radiation damage is controlled by a thermally 

activated processes, so some temperature 

control is desirable 

• Planned Work 

– Neutronics/gamma heating calculations 

– Thermal hydraulics calculations (safety case) 

– Can design and T&C specifications 

Muroga (2001) 

OPAL contains empty space (“hot 

source”) in the RPV. By developing a 

suitable holder (such as the ORNL 

HFIR sample heating holder, left), 

samples may be irradiated in OPAL at 

higher temperatures, thereby enabling 

elevated-temperature damage mechanisms 

to be characterised.

Effect of water ingress 

• XRD shows the formation of hydroxide species in the salt 

– Potassium hydroxide, sodium hydroxide and hydrated versions of these 

• Solid state nuclear magnetic resonance spectroscopy (NMR) 

– Exploits magnetic properties of certain nuclei 

• Nuclei absorb electromagnetic radiation and signals shift based on bonding environment 

(hydrogen and fluorine are good nuclei – allows for quantitative determinations) 

• Hydroxide and other species measured by proton NMR, Fluorine NMR used to measure main 

slat components which show good agreement to the literature 

– Extra peaks associated with Na are due to higher resolution instrument and larger spectral window 

X-ray diffraction 

19 

F Solid-state nuclear magnetic 

resonance spectroscopy

Effect of water ingress 

• Vibrational spectroscopy allows the 

chemical speciation to be determined 

– Raman spectroscopy measures bond 

polarisation 

– Infrared measures dipole moments 

• Observe the formation of HF species 

in the salt 

– Associated with KF (extremely 

hygroscopic and deliquescent)

H 2 O Decomposition in Fluoride Salts 

• Molecular dynamics using DFT used to study the stability of H 2 O in fluoride salts. 

• Deviation from stoichiometry found to break H 2 O into OH - and free H + (forming 

HF in hyper-stoichiometric salt). 

• Calculations carried out at 1000 K with a time step of 1 fs. (Run for 0.1 ns). 

• Phenomenon seen in non-stoichiometric F 2 LiNa and F 3 LiNaK. 

Stoichiometric F 3 LiNaK 

Non-Stoichiometric F 3 LiNaK

FE Model 

- Thermo-Mechanical 

Analysis 

- 3D Half Model 

- 407222 (HEX, 

Quadratic) 

- Thermal Analysis: 

DC3D20 

- Mechanical 

Analysis: C3D20R 

76 mm (Z) 

200 mm 

(Z) 

► NeT TG6 three-pass Inconel 82 slot-weld in 

an Inconel 600 (international benchmark 

specimen). The Abaqus half model depicting 

the basic plate geometry 01 and three 

150 mm (X)

Material Properties 

► Comparison of the room temperature cyclic responses of Inconel 600 (parent metal) and Inconel 82 (weld metal) predicted by 

mixed isotropic-kinematic hardening plasticity model when using the first and second cycle to derive mixed hardening 

parameters. 

02

WRS - FE Modelling 

Y 

X 

Z 

D Plane 

04

WRS - FE Modelling 

Y 

X 

Z 

B Plane 

04

ND Measurement 

150 mm (X) 

200 mm 

(Y) 

- Monochromator: bent Si single crystal 

- Monochromator angle: 33.52 

- Take-off angle: 66.98° 

- Detector angle: 88.0° 

- Wavelength: 1.498Å 

- Measured reflection: {311} 

- Primary slit: 2mm 

- Secondary slit: radial collimator 

- Gauge volume: 2x2x2mm 2 

ε 311 ij = d ij 

311 311 

− d 0,ij 

311 

d 0,ij 

Parent Metal 

Weld Metal 

σ 311 E 311 

ij = 

(1 + υ 311 ) ε ij + 

υ 311 

1 − 2υ 311 (ε 11 

311 + ε 311 22 + ε 311 33 ) 

Big Grains in 

the Weld 

Region 

05

d0 Measurements 

06

WRS - Modelling & Measurement 

Something seems to be 

wrong here. It looks that 

for the transverse 

direction we are B2 using Line 

the d0 from the parent 

B6 Line 

metal, while for the 

longitudinal direction we 

seems to be using the 

d0 from the weld metal. 

Can you please double 

check this. Something is 

certainly not right, 

because the model 

agrees in one direction 

and disagree in another 

direction. It should 

behave consistently 

between transverse and 

longitudinal. Similar to 

lines B2 and B10. 

07


B10 Line 

BD Line 

08


D2 Line 

D5 Line 

09


This doesn’t seem to be 

right. We are measuring 

lower stresses along 

D7.5 than along Line 

D10. And again the 

model agrees in two 

directions and disagree 

in one direction, that 

doesn’t seem to be right. 

For D7.5 line we should 

use the same d0s as for 

line D10. How did we get 

the d0s along the line 

D7.5 line? I don’t see a 

reason why it should be 

different from the D10. 

D7 Line 

D10 Line 

10

Admiral Hyman Rickover (US Navy) 

An academic reactor or reactor plant almost always 

has the following basic characteristics: 

(1) It is simple. 

(2) It is small. 

(3) It is cheap. 

(4) It is light. 

(5) It can be built very quickly. 

(6) It is very flexible in purpose. 

(7) Very little development will be required. It will 

use off-the-shelf components. 

(8) The reactor is in the study phase. It is not being 

built now.


On the other hand a practical reactor can be 

distinguished by the following characteristics: 

(1) It is being built now. 

(2) It is behind schedule. 

(3) It requires an immense amount of development 

on apparently trivial items. 

(4) It is very expensive. 

(5) It takes a long time to build because of its 

engineering development problems. 

(6) It is large. 

(7) It is heavy. 

(8) It is complicated.


• The tools of the academic designer are a piece of paper and a 

pencil with an eraser. If a mistake is made, it can always be 

erased and changed. 

• If the practical-reactor designer errs, he wears the mistake 

around his neck; it cannot be erased. Everyone sees it. 

• The academic-reactor designer is a dilettante. He has not had to 

assume any real responsibility in connection with his projects. 

He is free to luxuriate in elegant ideas, the practical 

shortcomings of which can be relegated to the category of "mere 

technical details.” 

• The practical-reactor designer must live with these same 

technical details. Although recalcitrant and awkward, they must 

be solved and cannot be put off until tomorrow. Their solution 

requires manpower, time and money. 

In Nuclear, materials are usually in this 

category of “mere technical details”

Overview of R&D Activities at ANSTO 

Cory J Hamelin 

Australian Nuclear Science and Technology Organisation, New Illawarra Road, Lucas Heights NSW 2234 Australia 

Presented at the 13 th Meeting of the Materials Project Management Board for the GIF VHTR System 

Jeju, Republic of Korea, 7-9 September 2015

74

ANSTO and Nuclear Fuel Cycle R&D 

• Fuel/cladding interactions 

– Use of atomistic modelling (e.g. DFT) 

• Structural materials performance under extreme 

conditions 

– Irradiation, corrosion, high temperatures and/or deformation 

• Separation science 

– Synthesis and analysis of titania frameworks for separation of U, Pu 

• Wasteform fabrication 

– Construction of a Synroc waste treatment plant to reduce volume of ANSTO 

nuclear by-products by 99% 

Overview of ANSTO R&D Activities 

13th Meeting of the Materials Project Management Board for the GIF VHTR System 

75

ANSTO and Generation IV Structural Materials R&D 





• Corrosion 





– Irradiation under high temperature 



76






• Corrosion 








77

Radiation Damage Test Environments at ANSTO 

• 1MV VEGA accelerator 

• 2MV STAR tandetron accelerator 

• 6MV SIRIUS tandem accelerator 

• 10MV ANTARES tandem accelerator 

– Energies from < 1 MeV to 100 MeV 



78

79

Specimens for Neutron Irradiation 

• Miniature dogbone samples 

• Small punch discs 

• Compact tension discs 

• Quarter-size Charpy samples 

• Corrosion samples 

1 dpa/yr 



80

Samples Currently in OPAL (or Planned for Insertion) 

Material Sample type Irradiation time Source 

MA957 TEM discs 460 days IME/ORNL i 

CLAM “ “ IME i 

TiAl FT “ “ IME/LANL ii 

AlMg5 Mini tensile “ IME/OPAL Surveillance iv 

Zr-3.5Sn “ 460 days IME/Queens ii 

TiAl DP TEM discs 460 days IME/LANL ii 

Ti-6Al-4V “ “ IME i 

Zr-2.5Nb “ “ IME i 

Zr-4 CT 460 days OPAL Surveillance iv 



NiCrMoFe Small Punch 460 days IME/SINAP JRC iii 


Ti45Al “ 460 days IME i 


Ti45Al TEM disc “ IME i 


Zr-2.5Nb “ “ IME/OPAL i 

NiCrMoFe “ “ IME/SINAP JRC iii 

Zr-3.5Sn Mini Tensile 5 years IME/Queens ii 


Al6061 “ 5 years OPAL Surveillance iv 

i 

Internal research 

ii 

unfunded collaborative research 

iii 

partially funded research with external partners 

iv 

OPAL Materials Surveillance Program (MSP) 

• Zr and Al alloys 

– Predominantly for OPAL support 

• Titanium Aluminides 

– Internal research 

• Ni alloys 

– Collaborative work (SINAP MSR) 

• Additional planned materials 

– A508 (PM, WM, HIP) 

– ODS 

– Graphite 

– Tungsten 



81

Active Sample Machining and Testing 

• Active samples sent to MEHC, which contains: 

– Mechanical testing facilities for active material 

– Micromachining facilities to allow samples to exit cells 

• MEHC currently in licensing/commissioning stage 



82

The Role of Ion Irradiation 

• Current limitations on neutron fluence 

– Irradiating at coolant water temperature (higher temperatures desirable) 

• Ion irradiation allows for higher dpa 

– Proton and alpha damage of thin samples 

– Self irradiation of energetic heavier ions 

• Must be accompanied by modelling of the processes 

to enable correlation of ion vs. neutron damage 

• Useful for preliminary qualification of nuclear structural 

materials 



83

Single Crystal Ni, He+ Ion Irradiation 

11.85 µm 

~ 19 dpa 

Irradiation 

direction 

Front surface 

(Entry surface) 

Back surface 

(Exit surface) 

~ 10 dpa 

SRIM (The Stopping and Range of Ions in Matter) estimates for He+ irradiation of Ni 



84

Micromechanical Testing 

Radiation direction 



85

Trends in Irradiation Strengthening with Dose 

• Materials performance under ion irradiation provides useful qualitative information 

• Full-sample performance does not explicitly consider localised damage 



86

Layered Irradiation Studies 

Ar implanted layer 

Ar implanted layer 

He implanted 

layer 

Cu tape 

• Increases width of peak damage layer in specimen 

– Subjected to hardness testing to determine deformation mechanism(s) 

• Materials tested include Ti6Al4V, ODS (MA957), Zr-4 

and AISI 316 



87

Variation in Radiation Damage Morphology 

Cross-section TEM of irradiated and indented AISI 316 sample (extracted with FIB) 

Top surface 

Few bubbles 

(Ar) 

Radiation damage 

Ar 

He 

Bubbles 

(He) 

Bubbles 

(He) 



88






• Corrosion 








89

Creep Rupture Assessment 

• Member of the R5 EDF Energy Code of Practice development panel (“Assessment 

Procedure for the High Temperature Response of Structures”) 

• Component life assessment using advanced damage accumulation models (SMDE, 

SEDE), considering multi-axial effects 

• Materials studied largely ferritic steels (P22, P91, CMV, etc.), some austenitic steels 

(e.g. AISI 316H) and Ni alloys (Ni 201, Inconel 600) 

1000 

450 C 

Stress (MPa) 

500 

450 

400 

350 

300 

250 

200 

150 

ANSTO 

NIMS 

JAEA 

CFAT 

CRIEPA 

ORNL 

NIMS-2 

ORNL2 

Energy Density MJ/m 3 

100 

10 

500 C 

550 C 

600 C 

650 C 

700 C 

Fit 450 C 

Fit 500 C 

Fit 550 C 

Fit 600 C 

Fit 650 C 

100 

ODIN 

Fit 700 C 

50 

0 

1 10 100 1000 10000 100000 

Time to Rupture (Hours) 

1 

0.000001 0.00001 0.0001 0.001 0.01 0.1 1 

Average Strain Rate (1/hr) 

Creep rupture data for P91, taken from a variety of sources (left). Energy density predictions using hybrid energy models (right). 



90

Predicting Creep Rupture 

• FEA prediction of in-service materials performance 

• User-defined material models for creep strain/damage 

Prediction (via FEA) of creep rupture during an accelerated creep test in ex-service AISI 316H stainless steel using 

a Strain Energy Ductility Exhaustion (SEDE) creep damage model 



91

Quantification of Accumulated Creep via EBSD 

• Data analysis for performance trends 

– Quantifying plasticity (creep vs. fatigue) via EBSD 

– Studies on Ni-201 and P22 (underway) 

Example of damage characterisation in Ni-201 material: (left) virgin material; (centre) material subjected to 

15% creep strain; and (right) damage curve constructed to assess creep life 



92

Creep-Fatigue Tests 

• Laboratory testing predominantly high-Cr ferritic steels (P91, P92) as part of 

collaborative study in MATTER FP7 (MATerials Testing and Rules for Generation IV 

Reactors), WP4 (Prenormative R&D for Codes and Standards) 

Observed Nf 

100000 

10000 

1:1 

X2 

X2 

Energy Modified Exhaustion 

Stress Modified Exhaustion 

RCC-MR (Time Fraction) 

Time Fraction 

Ductility (Elongation) 

Power (Energy Modified Exhaustion) 

Power (Stress Modified Exhaustion) 

Power (RCC-MR (Time Fraction)) 

Power (Time Fraction) 

Power (Ductility (Elongation)) 

1000 

100 

10 100 1000 10000 100000 

Predicted Nf 

ANSTO P91 creep-fatigue tests were used with test data from ORNL, CEA, EPRI and IGCAR to assess the accuracy of new and existing 

damage models (TF, SMDE and SEDE shown) in predicting material failure. Test data conducted at 450-700 ºC, resulting in rupture times 

ranging from 2-50,000 h. 



93

P91 Japan Sodium-Cooled Fast Reactor Tubesheet 

• Hot transient sodium temp = 600 C (2hr dwell) 

• Cold transient temp = 250 C (1 hr hold) 

• Rate Change = 5 C/s 

• Cycle length = 3 h and 140s per cycle 

• Number of test cycles = 1873 

• Crack initiation mode creep-fatigue 

• Maximum Crack Length = 3.7 mm 

Ando (2014) 

Original JAEA analyses performed using RCC-MR 

time fraction (TF) models for P91. Strain ranges 

derived using: (i) Stress Redistribution Locus, SRL; 

(ii) Simple Elastic Follow-Up, SEF; and (iii) direct 

inelastic implementation using FEA. ANSTO-EDF 

Energy analyses (in red) used inelastic strain range 

with ductility exhaustion models. 



94

Creep-Fatigue Materials Database 

• Creation of a materials database for creep-fatigue 

performance 

Full C-F Model 

1Cr0.5Mo (P12) 1.25Cr0.5Mo (P11) 2.25Cr1Mo (P22) 

0.5Cr0.5Mo0.25V (CMV) 1Cr1Mo0.25V (CMV) 1.25Cr1Mo0.25V (CMV) 

P9 X10CrMoVNb (P91) 9Cr0.5Mo1.8WVNbB (P92) 

AISI 304 AISI 316 25Cr35NiNbMa (HK40) 

X20CrMoV12-1 7-9Cr2WV P23 

Hastelloy XR Fe0.75Ni0.5MoCrV Fe2.25Cr1Mo0.25V 

Under Development 



95






• Corrosion 








96

Corrosion Testing 

• ANSTO-SINAP JRC (MSR) 

– FLiNaK salt composition 

– Tests conducted at 750ºC for 10, 100 

and 200 h 

• Materials tested 

– NiCrMoFe 

– Graphite 

– AISI 316 

Controlled atmosphere furnace (above) with temperature 

control directly linked to salt temperature; exposed NiCrMoFe 

is examined via EDS (left) to determine surface degradation. 



97

Salt Infiltration into Graphite 

Salt infiltration into various grades of graphite (right, top) at different pressures 

(above): (a) 1.0 atm; (b) 1.5 atm; (c) 3.0 atm; and (d) 5.0 atm. Location of salt 

quantified via EDS (right) and neutron tomography (DINGO). 



98






• Corrosion 








99

Combined Effects of Corrosion and Ion Irradiation 

• NiCrMoFe alloy, FLiNaK salt, 10 17 ions/cm 2 He+ 

As-Received 




Pt-Ni interface measured using STEM-EDS under four conditions representing various combinations of corrosion and radiation damage. The 

relative increase in interface layer before and after corrosion represents the corrosion layer. This layer is 20 nm for the as-received material, 

and 700 nm for the irradiated material. 



100

Combined Effects of Corrosion and Ion Irradiation 

• Graphite, FLiNaK salt, 10 17 ions/cm 2 He+ 

As-Received 




Graphite test coupons were irradiated using 30 keV He+ ions; one section of the 

sample was masked to prevent radiation damage (right). The samples were 

then exposed to molten FLiNaK salt (150 h @ 750 ºC). Clear variations in the 

corroded structures are visible (above). 

Damaged 

Masked 



101






• Corrosion 








102

Materials Testing at High Temperatures 

• NiCrMoFe creep rupture test (190 MPa @ 700 ºC) 



103






• Corrosion 








104

High Temperature Irradiation (Planned) 

• Motivation 

– Radiation damage is controlled by a thermally 

activated processes, so some temperature 

control is desirable 

• Planned Work 

– Neutronics/gamma heating calculations 

– Thermal hydraulics calculations (safety case) 

– Can design and T&C specifications 

Muroga (2001) 

OPAL contains empty space (“hot 

source”) in the RPV. By developing a 

suitable holder (such as the ORNL 

HFIR sample heating holder, left), 

samples may be irradiated in OPAL at 

higher temperatures, thereby enabling 

elevated-temperature damage mechanisms 

to be characterised. Kiritani (1988) 



105

Additional Considerations: Welding 

• Understanding of microstructure and stress state 

around single- and multi-pass welded joints 

• Heavily involved in international round-robin 

programmes designed to identify best practise for joint 

characterisation and process simulation 

– NeT (AISI 316, Inconel 600, A508) 

– USNRC (DMW) 

• ANSTO carries out materials testing for materials data 

set required for simulation, as well as characterisation 

and performance evaluation of welded joints 



106

Material Test Data (High-Temperature LCF) 



107

Process Simulation: DMW 



108

Process Simulation: EB Weld (A508) 

Temperature 

Distribution 

Stress 

Distribution 

Martensite 

Formation 

Bainite 

Formation 

Courtesy A. Vasileiou, University of Manchester 



109

ANSTO-University of Manchester RCA 



110

Weld Characterisation (NiCrMoFe) 



111

Conclusions 

• Variety of structural materials R&D has been, is and will 

be conducted at ANSTO 

– Much of that work is related to Generation IV systems, particularly for VHTR and 

MSR environments 

• Radiation, high-temperature, and corrosion 

environments considered (alone or in concert) 

– Characterisation (both ex situ and in situ), testing and analysis performed 

• Numerous materials under investigation 

– Apart from materials critical to ANSTO infrastructure, feedback through bodies 

such as the VHTR PMB essential for future research planning 



112

ANSTO as a National Laboratory 

• Undertake and facilitate research in the national 

interest 

• Produce and provide state-of-the-art 

radiopharmaceuticals for Australia and the world 

• Provide trusted advice to Government in all 

aspects of the Nuclear Fuel Cycle 

• Operating ANSTO’s nuclear facilities including the 

OPAL nuclear research reactor in a safe and 

efficient manner

ANSTO Nuclear Medicine Project 

• 80% of all nuclear medicine procedures use 

Mo-99/Tc-99m 

• Diagnosing heart disease, cancer, 

neurological disorders and more. 


• 40 million patients per year 

• Global market of US$550 million per annum 

• Potential critical worldwide Mo-99 shortage 

• A new Australian Mo-99 production facility 

• Mo-99 production capacity to allow for 

increased exports 

• First full scale operational Synroc facility in 

the world

GH3535 Creep Conclusions 

• GH3535 alloy creep tested at 650, 700, and 750 o C at 85-320 MPa. 

• Alloy displayed no primary creep and good creep ductility (above 30%.) 

• Using minimum creep rates the calculated stress exponent is n = 5-6 so 

dislocation climb is the main creep mechanism 

• Formation of dislocation networks and sub-grains confirmed by TEM and 

EBSD analysis. 

• The Dorn Shepard and Larson Miller master curves were derived to 1% 

creep strain and creep rupture respectively. 

• Using ASME BPVC guidelines the maximum allowable design stress at 

700 o C is 35 MPa. 

• There is substantial second phase particle precipitation during creep.

Australian’s Molten Salt Reactor (MSR) Material Research Ondrej Muránsky

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