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Copyright © 2015
SCK•CEN
Copyright © 2015
SCK•CEN
R&D in Fusion Materials
Dmitry Terentyev Fusion Project Manager
VeMet meeting 19/06/2015 Mol, Belgium
Copyright © 2015
SCK•CEN
SCK•CEN in short
Cradle of Belgian nuclear research, applications and energy
development in Belgium
Major international player in the field of nuclear R&D
~700 staff, >50% with academic degree + 70 PhD students
Annual turnover: 140 M€
45% government support
55% contract work
Copyright © 2015
SCK•CEN
28 european partners
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SCK•CEN
SCK•CEN Organisation
General Management
Institute
Nuclear Materials
Science
NMS
Institute
Advanced Nuclear
Systems
ANS
Institute
Environment, Health
& Safety
EHS
Institute
Corporate Services
and Administration
CSA
MYRRHA
Management Team Communication
Safety Business Support
A
C
A
D
E
M
Y F U S I O N
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SCK•CEN
General lines of R&D of
Nuclear Material Science Institute
Institute of Nuclear Materials / Structural Materials Group
Copyright © 2015
SCK•CEN
Structural materials group: RESEARCH TOPICS
1) Reactor Pressure Vessel Steel - Surveillance, Material characterisation
- Ageing, irradiation damage & embrittlement modelling 2) Austenitic and Ni based alloys
- Corrosion susceptibility testing; focus on IASCC -µ-structure & electrochemical behaviour
Copyright © 2015
SCK•CEN
Structural materials group: RESEARCH TOPICS
3) GEN IV structural materials
- Compatibility with liquid metal - Irradiation damage and embrittlement
4) Fusion structural materials - 14 MeV hardening & He-embrittlement - Plasma wall interaction
Copyright © 2015
SCK•CEN
fusion technology R&D programme
The materials development
Neutron effects on the material properties;
Based on large and specialized infrastructure: MTR, hot cells, theoretical support …
Long tradition and broad skills, also developed in fission
Focus on PFM (W) and structural materials (9Cr)
The behavior of components in radiation fields (e.g. diagnostics & FOCS)
Focused on optical instrumentation and remote handling components
Development of radiation hardened components
Possibilities of irradiation and on-line measurements
Copyright © 2015
SCK•CEN
fusion technology R&D programme
The waste management and tritium handling
Extended experience in radioactive waste
management, effluents handling and R&D on
decontamination
Refurbishment of the tritium lab with increased
tritium content license (mostly for the plasmatron)
Combination of tritium handling and PWI (tritiated
plasma capacity)
Other topics also covered:
Irradiation device design and development (for
IFMIF e.g. but also for the material development,
testing and qualification)
Socio-economics studies, based on experience in
the fission domain
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SCK•CEN
Neutron Irradiation of materials and components
Charpy Reconstitution
Milling Machine
Preparation for metallography
TEM disc slicing
Sample irradiation
Sample preparation
Hot cell operation
BR2
Hyperboloid of Revolution 1015 n/cm² thermal 8x1014 n/cm² fast n-flux 0,5 dpa/cycle
Copyright © 2015
SCK•CEN
Gamma irradiation of diagnostics and components
Brigitte
( 60 Co - fuel) RITA
( 60 Co)
Geuse II
(fuel)
LNC
( 60 Co)
Dose - rate
max.
Dose - rate
min.
Vol. (mm 2 )
Vol. Temp.
VUB
Cyclotron
1.4 krad•s - 1
140 rad•s - 1
300 rad•s - 1
30 mrad•s - 1 2 rad•s - 1
15 rad•s - 1 140 mrad•s - 1
900 x 220
900x80
600 x 380 400 x 380
50 - 200 °C RT - 100 °C RT RT stabilised
BR1
BR2
Hot cell
0.3 mrad•s - 1
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SCK•CEN
Post-irradiation examination in Hot Cells
SEM
OM
EPMA
XRD
XPS
TEM
PA
MAE IF
Tritium lab
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SCK•CEN
In-pile creep testing
In-pile fatigue specimen
Copyright © 2015
SCK•CEN
Material testing under extreme conditions
- Creep, swelling and He release during long term annealing - Tensile tests and other characteristics at high temperature (W) up to 2000°C !
Copyright © 2015
SCK•CEN
Specific device: Plasma Wall Interaction
Plasmatron installation, completely
refurbished
Work on radioactive samples, on Be
and with tritium gas
unique facility in Europe
First plasma: Jan. 2009
Start experimental tests from mid 2010,
stepwise with H and then D.
Already applied: D-retention in W
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SCK•CEN
Fusion Material Research at SCK•CEN:
Embrittlement of high-Cr steels
Institute of Nuclear Materials / Structural Materials Group
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SCK•CEN
Fusion reactor structural materials
ITER
Expected
dose: 3 dpa
Austenitic stainless
steel 316L
Beryllium (Be) -
1st wall
Tungsten (W) -
Divertor
Design is established
Choice of structural materials based on extensive tests performed in the past
Suitable for low dose
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SCK•CEN
Fusion reactor structural materials
DEMO
Expected
dose: ~150-
200 dpa
No established design
Different possible choices of structural materials depending mainly on desired operation temperature
Vacuum
vessel
Breeding
blanket
Divertor
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SCK•CEN
0
0,1
0,2
0,3
0,4
0,5
0,6
0,7
0,8
0,9
1
0,1 1 10 100 1000
Radiation effects in structural materials
Dose received (dpa)
Ho
mo
log
ical
Tem
pera
ture
(T
/TM
)
> 0.1 dpa, <0.35 TM
Radiation hardening and embrittlement
>> 10 dpa, T>0. 5 TM
If He>100appm
He embrittlement at GB
(intergranular fracture)
>10 dpa, 0.3TM<T<0.6TM
Phase instabilities from
radiation-induced
segregation and
precipitation
Volumetric void swelling
(dimensional instability)
>10 dpa, 0.35TM<T<0.45TM
Irradiation creep
Copyright © 2015
SCK•CEN
Getting engineering data
Full scale Charpy sample: Tensile, Creep, 3P/4P bending, Fracture toughness
Copyright © 2015
SCK•CEN 22
Physical origin of hardening Clean material Irradiated material
Hardening = increase of the yield stress … after treatment such as : - thermal annealing - deformation - irradiation
Displacement
Lo
ad
Baseline
Irradiated
04 Oct 2001 29 Nov 2001
Copyright © 2015
SCK•CEN
APril 2014 – D. Terentyev – Radiation Damage in Fusion Materials 23
Brittle behaviour Brittle material = material breaks prior plastic deformation
Embrittlement = 1. reduction of elongation/deformation
before fracture
2. increase of temperature below which
material is brittle
T
c
y DBTT
Definition of ductile-
brittle transition
temperature (DBTT, or
Tc):
Yield stress = Cleavage
stress
Str
ess
Strain
y
10mm
Copyright © 2015
SCK•CEN 24
Nanostructural changes macroscopic changes Brittle material = material breaks prior plastic deformation: why ?
Unirr. Str
ess
Strain
Yield point
T < Tc
> c
T > Tc
< c
dislocations
emitted before c
reached ductile
obstacles to
dislocation emission
= embrittlement
brittle behaviour
Because the origin of hardening and embrittlement is the presence of radiation
defects obstructing dislocation motion
hardening and embrittlement are generally correlated!
Copyright © 2015
SCK•CEN
APril 2014 – D. Terentyev – Radiation Damage in Fusion Materials 25
DBTT shift correlates wtih the yield strength change
Displacement
Lo
ad
Baseline
Irradiated
04 Oct 2001 29 Nov 2001
yield increase
400
600
800
1000
1200
-200 -100 0 100 200 300
temperature (°C)
yie
ld s
tres
s (M
Pa)
baseline
irradiated
yield increaseyield increase
Radiation defects obstruct dislocation movement
Copyright © 2015
SCK•CEN
APril 2014 – D. Terentyev – Radiation Damage in Fusion Materials 26
T
c
y
DBTTH
y,irr
Irradiation hardening
DBTT
Irradiation
embrittlement
DBTT shift correlates wtih the yield strength change
Copyright © 2015
SCK•CEN
DBTT and y are generally linearly correlated
y
D
BT
T
Empirical correlation
To = 0.3
y RAFM
To = 0.7
y RPV
YIELD STRENGTH INCREASE, MPa
0 100 200 300 400 500 600
T
o S
HIF
T, o
C
0
50
100
150
200
250
F82H-IEA
F82H-HT2
9Cr-2WVTa
Eurofer97(Lucon, SCK-CEN)
Eurofer97(Rensman, NRG)
RPV Steels
(Sokolov, ASTM STP 1325)
Because the origin of both hardening and embrittlement is
the presence of obstacles to dislocation motion
hardening and embrittlement are generally correlated!
But different materials exhibit different correlation factor
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SCK•CEN
Engineering modelling: hardening
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SCK•CEN
Engineering modelling: microstructure
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SCK•CEN
Engineering modelling: accounting for He
Copyright © 2015
SCK•CEN
Engineering modelling: model calibration
Copyright © 2015
SCK•CEN
Engineering modelling: model application
0.5 – 1 dpa … is where engineering
modelling breaks … and physical
modelling is needed
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SCK•CEN
Experimental Approach:
Prepare and
characterize material
Irradiate material in
reaction
Characterize µ-structure
changes
Check mechanical
properties
In-pile response is different from the out-pile test
Need to fill gaps, where the irradiation conditions are out of experimental accessibility
In-reactor tests (in BR2): S. Tähtinen (VTT) and B. Singh (DTU)
Δσ
dose DBTT
Need for Material’s Modelling
Copyright © 2015
SCK•CEN
Modelling framework
Primary damage : 10-9 seconds, 10-8 meters:
1. creation of defects and small defect cluster
2. products of transmutation
Short term : radiation induced-diffusion: 10-3 seconds, µm:
1. formation of dislocation loops and nano-voids
2. Segregation and precipitation of alloying elements
Ageing : reorganization of lattice defects: 10-3sec –years, grain
1. Formation of He bubbles and void lattices
2. Growth and coalescence of precipitates (M-C), change of GB composition
3. Formation of dislocation network and forest
4. Phase decomposition
Mechanical tests: seconds-hours, single-poly crystals:
1. Dislocation-defect interaction
2. Trans/inter granular embrittlement
n
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SCK•CEN
Contribution of Material’s Modelling
Primary damage state
Survived defects
Damage morphology
Fission vs. Fusion
Evolution of microstructure
Accumulation of damage
Chemical changes
“Invisible” damage
Plastic deformation
Localized deformation
Fast deformation
-120 -90 -60 -30 0
-5
0
5
10
15
20
F - resisting forceF - resisting force
2R -obstacle size
DPRP
=3.5 nm
dislocation line
[11
1]
(a0)
[11-2] (a0)[1-10]
L -obstacle spaing
- applied stress
T - line tension
Copyright © 2015
SCK•CEN
Contribution of Material’s Modelling
0.0
25
µm
0.02 µm 0.02 µm
MD DD
100 200 300 400 500 600 700 800 9000.0
0.2
0.4
0.6
0.8
1.0
m
ax (
Gb/L
)
DD results
MD results
Temperature (K)
DL = 5 nm
VD = 20 m/s
Hardening and plasticity at meso-scale
Taking into account realistic defect distribution
Effects of local strains and temperature excursion
Addressing experimentally “invisible” defects
4 M atoms
100 CPU days
1 K segments
0.0001 CPU days
Copyright © 2015
SCK•CEN
Contribution of Material’s Modelling
1E-3 0.01 0.1 1
0
50
100
150
200
250
300
Dose (dpa)
(
MP
a)
Experiment
Voids
Loops
Loops + Voids
5 µm
5 µ
m
Heterogeneous plastic deformation
Mechanisms of channeling
Conditions favouring channel nucleation
Copyright © 2015
SCK•CEN
Evolution of material’s modelling over last 50 years
First experimental observations 1955-1962 (UK)
Swelling predicted in 1959, and discovered in 1966
RIS predicted in 1972, and discovered in 1973
Atomistic simulations 1964 – till now
1964 – 1986: simple pair interaction physics
1986 – 2000: large scale physics: dislocations, grain boundaries
From 2000: first principle calculations: electronic and chemical effects
1962 2013
Consistent explanation of µ-structure in Fe and Fe-C under neutron irradiation at T ≤ 300°C
2000 1986
Observations
and hypothesis
Classical MD In situ & Ab initio Hybrid techniques at
meso-scale
5 µm
5 µ
m
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SCK•CEN
Irradiation using BR2 reactor for fusion
needs R&D programme
Institute of Nuclear Materials / Structural Materials Group
Copyright © 2015
SCK•CEN
Reconstitution Technology & Specimen Miniaturization
Miniature
tensile Charpy specimen
Reconstitution
Broken Charpy specimen
Reconstitution Technology
Other possibilities
microstructural samples
hardness samples
re-irradiation
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SCK•CEN
Historic perspective of BR2 Construction and commissioning period
1956: start of BR1
Construction: 1957-1960
First criticality 06/07/1961
Commissioning: 1961-1962
1962: first criticality of BR3
Initial operation license for 25 years issued in 1963
Copyright © 2015
SCK•CEN
What is the BR2 reactor?
The BR2 reactor is a materials test reactor
Purpose: production of neutrons for research
Compact reactor core: 30 times smaller than power reactor
Low temperature and energy: 40°C vs 300°C and 30 times less thermal output than power reactor
High neutron density: 20 times higher than power reactor
The BR2 consists of:
Reactor vessel inside reactor pool
Primary cooling loop
Secondary cooling loop
Reactor and machine buildings
Cooling towers
Auxiliary systems (ventilation, back up power supplies…)
Copyright © 2015
SCK•CEN
Individual irradiation
channels
Flexible
configuration of
reactor core
Maximal density in
center and maximal
accessibility at
extremity of
channels
Access from reactor
cover and sub-pile
room possible
The reactor interior
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SCK•CEN
Flexibility in applications: variable core configuration
Flexible to accommodate
experimental needs
Fuel elements
Control rods
beryllium matrix
Isotope production
CALLISTO (PWR simulation)
SIDONIE (irradiation of Si for
doping)
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SCK•CEN
Reactor building: containment of reactor and experimental devices
Copyright © 2015
SCK•CEN
Utilization of BR2
Material test reactor (MTR):
Goal: expose materials to neutrons in controlled conditions
Requirements:
Flexibility to create and combine different irradiation conditions
Access to irradiation channels, eventually also during irradiation
Access for on-line instrumentation and separate cooling loops
Applications
Testing of fuels for different reactor types
Accelerated irradiation of materials to predict degradation in service in
power reactors
Modification of materials for non energetic applications: production of
radio-isotopes and semi-conductors
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SCK•CEN
Capsule irradiations
Materials exposed to primary water flow or pool water
Loadable in reflector channel, fuel element, thimble tube.
Temperature dependent on capsule design, control by adjustable coolant flow possible.
Gas capsule: irradiation temperature determined by design and irradiation position – Up to 1000°C
Liquid metal capsule: irradiation temperature controlled by inter-wall gas volume – Up to 600°C
Pressurised water capsule: irradiation temperature controlled by water pressure (boiling point) – Up to 300°C
Irradiation time flexible.
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SCK•CEN
Thimble tube capsule holder “ROBIN”
Specimen holder geometry with gas gap
and metal matrix sample holder
Incorporation in closed needle
Monitoring by measurement of
temperature in dummy specimen
Temperature control by water flow
adjustment
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SCK•CEN
Thimble tube capsule holder “LIBERTY”
Common design IMR – SCK•CEN
Specimen holder geometry with gas gap and
metal matrix sample holder
Incorporation in closed needle
Temperature control by adjustment of the
electrical heater power
Temperature range from 50 to 1000°C
Maximum specimen cross section = 10 x 10
mm²
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SCK•CEN
Irradiation in “LIBERTY”
Holder preparation
Specimens
Needle
Loading the needle
in “LIBERTY”
Loading “LIBERTY”
in the thimble
Temperature Control
Copyright © 2015
SCK•CEN
Boiling water capsule irradiation “MISTRAL”
200-300°C irradiation temperature
Boiling water environment
High Fast Flux level
Irradiation temperature monitoring and
control by water pressure and heating
element
Reloadable with standard specimens
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SCK•CEN
Simulation of PWR in BR2: the CALLISTO loop
Full thermal-hydraulic
simulation of PWR
conditions
Independent cooling
system
Fuel: 3 x 3 assembly
1m rods
Structure materials
Used for Fuels and Structural Materials Irradiations
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SCK•CEN
Pb-Bi capsule inside a fuel element “SPEED ASTIR”
450°C in lead-bismuth eutectic
Highest possible fast flux
In a standard six-plate fuel element
2.5 to 3 dpa in 5 cycles
Temperature monitoring + control
Two heating elements
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SCK•CEN
MYRRHA
Accelerator (600 MeV – 2.5 mA proton)
Fast
neutron
source
Spallation source
Lead-Bismuth
coolant
Multipurpose flexible
Irradiation facility
Reactor • subcritical mode (50-100 MWth)
• critical mode (~100 MWth)
SCK•CEN builds the
innovative
international
research reactor
MYRRHA in Mol
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SCK•CEN
Copyright © 2015 - SCKCEN
PLEASE NOTE!
This presentation contains data, information and formats for dedicated use ONLY and may not be copied,
distributed or cited without the explicit permission of the SCK•CEN. If this has been obtained, please reference it
as a “personal communication. By courtesy of SCK•CEN”.
SCK•CEN
Studiecentrum voor Kernenergie
Centre d'Etude de l'Energie Nucléaire
Belgian Nuclear Research Centre
Stichting van Openbaar Nut
Fondation d'Utilité Publique
Foundation of Public Utility
Registered Office: Avenue Herrmann-Debrouxlaan 40 – BE-1160 BRUSSELS
Operational Office: Boeretang 200 – BE-2400 MOL