Post on 05-May-2018
EMIR - Orsay - 21 october 2011 Nuclear Materials Department Ph. Dubuisson - 1
Challenge of materials for nuclear reactors
fission and fusion
Ph. Dubuisson, P. Yvon
Orsay – France 21 october 201 1
EMIR Users days 20 – 21 october 201 1
Nuclear Materials Department
EMIR - Orsay - 21 october 2011 Nuclear Materials Department Ph. Dubuisson - 2
Outline
Requirements for nuclear materials non fissile Gen II-III
Gen IV
Fusion – Accelerator Driven System
Conclusions
EMIR - Orsay - 21 october 2011 Nuclear Materials Department Ph. Dubuisson - 3
Requirements for materials 40 - 60 years life time – Vessel, internals,…
Fast neutron damage Evolution of the microstructure
• phase instability, precipitation, voids, amorphization, … Dimensionnal changes swelling, growth, irradiation & thermal creep Modification of mechanical properties
• YS, UTS, elongation, toughness, … hardening, embrittlement, … Thermal creep time to rupture hydrogen & helium Embrittlement
Resistance at high temperature Mechanical properties YS, UTS, elongation, toughness, … Embrittlement at high temperature Thermal creep time to rupture, deformation Creep - fatigue interaction
Compatibility with the different environments primary and secondary fluid, fuel, reprocessing, … heat Exchangers Fuel Cladding Chemical interaction, Hydrogen cracking Corrosion & cracking by stress corrosion I- SCC, IASCC, …
point defects & clusters gas, transmutation
also incidental and accidental conditions
EMIR - Orsay - 21 october 2011 Nuclear Materials Department Ph. Dubuisson - 4
Complementary considerations
Availability and cost of materials
Fabricability, joining technology welding, … Low activity Maintenance & repair - waste
Inspection in service? Non destructive examination techniques
Safety approaches, licensing and qualification Codes and design methods RCC M, RCC MRx, …
R&D effort needed to establish or complement mechanical design rules and standards
Codification for the nuclear design specific Qualification of core materials
Decommissioning and waste management
Requirements for materials
EMIR - Orsay - 21 october 2011 Nuclear Materials Department Ph. Dubuisson - 5
GEN 2&3 - PWR - Irradiated Components
~ 300°C 0.1 dpa
40 60 years
300 – 400°C 10/15 dpa 5 – 6 years
300 – 380°C 30 - 120 dpa 40 60 years
neutrons temperature mechanical stresses environment
Core Internals Austenitic steels
Fuel Assemblies Zr alloys
Vessel Bainitic steel
1 6MND5 A508 Cl 3
Core Internals Nickel alloys
Control rods Austenitic steels
~ 320°C ~ 10 dpa few years
~ 320°C few 0.1 dpa
40 60 years
155 bars 293°C Water
H2, LiOH, B
328°C
Extension of lifetime
loops
EMIR - Orsay - 21 october 2011 Nuclear Materials Department Ph. Dubuisson - 6
PWR - Vessel steels
PWR Vessel bainitic steel 16 MND5 (A508 CI.3)
DBTT Shift Irradiation Decrease in USE Hardening Loss in ductility
0
50
100
150
200
250
-200 -100 0 100 200 300
Temperature °C
Toughness MPa √m
TT 0
50
100
150
200
250
-200 -100 0 100 200 300
0
50
100
150
200
250
-200 -100 0 100 200 300
∆TT
Start of life
ageing irradiation
Fluence en fin de viedes unités 900 MWe
(40 ans)
Fluence 10 n/cm (E>1MeV)19 2
∆TT
90 nm90 nm
VVER steels Cu P Mn Ni Si Clusters
GPM Rouen + point defect clusters
EMIR - Orsay - 21 october 2011 Nuclear Materials Department Ph. Dubuisson - 7
EtLD
Service
IPG
DENO APRP
RIA
Transport
0
100
200
300
400
500
0 500 1000 1500
Température (°C)
Co
ntr
ain
te (
MP
a)
lowveryT,ε
lowT,ε
highT,ε
highveryT,ε
PCI
RIA
LOCA
Service Dry-out
shipment storage
σ MPa
T °C
PWR cladding – Zr alloys
life time X2 in 15 years High density of <a> loop
Localization of deformation
EMIR - Orsay - 21 october 2011 Nuclear Materials Department Ph. Dubuisson - 8
0
200
400
600
800
1000
1200
1400
0 200Déplacement par atome (dpa)
Tem
péra
ture
(°C)
Générations II- III
GEN IV – 6 systems – Irradiation conditions
VHTR Gas Fast Reactor
Molten Salt Reactor
dose (dpa)
T (°C)
The most mature option
Lead Fast Reactor
Supercritical Water cooled Reactor
Sodium Fast Reactor
New goals for sustainable nuclear energy… New challenges for materials ! Here normal operating conditions
EMIR - Orsay - 21 october 2011 Nuclear Materials Department Ph. Dubuisson - 9
PHENIX
SPX1
SFR
PHENIX
SPX1
SFR
Dose > 150 dpa Stress 100 MPa Temperatures 400 – 650°C
8 – 10 years
SFR Cladding – Material Choice
ODS ferritic- Martensitic steels Nano dispersion
Improved safety - Reduce • Fuel enrichment • Reactivity excess • Potential void effect
Maximise fuel content reduce Na in the core
0
2
4
6
8
10
60 80 100 120 140 160 180 200
dose (dpa)
Average316 Ti
Ferr i t ic -m art ens it ic (F/M)st ee ls , ODS inc luded
Average 15/15Ti Best lot o f 15/15Ti(%)
Phénix
0
2
4
6
8
10
60 80 100 120 140 160 180 200
dose (dpa)
Average316 Ti
Ferr i t ic -m art ens it ic (F/M)st ee ls , ODS inc luded
Average 15/15Ti Best lot o f 15/15Ti(%)
Phénix
1
2
3
4
5
0,0001 0,001 0,01 0,1 1 10 100 1000 10000
650° C 1 80 MPa
ODS Fe-18Cr 0,5 Y2O3
Fe-18Cr
Time (h)
ε (%)
Low deformation Swelling, Irradiation Creep Thermal creep
EMIR - Orsay - 21 october 2011 Nuclear Materials Department Ph. Dubuisson - 10
PHENIX
SPX1
SFR
PHENIX
SPX1
SFR
Dose > 150 dpa Stress 100 MPa Temperatures 400 – 650°C
8 – 10 years
SFR Cladding – Material Choice Improved safety - Reduce
• Fuel enrichment • Reactivity excess • Potential void effect
Maximise fuel content reduce Na in the core
ODS Martensitic 9Cr steels ODS Ferritic 1 3 – 1 8 Cr steels
Nano dispersion
Low deformation Swelling, Irradiation Creep Thermal creep
Key "technological" issues Low deformation Swelling, Irradiation Creep Thermal creep
Elaboration Weldability
Mechanical prop. before, under and after irradiation toughness, DBTT, … embrittlement under irradiation
Behaviour in Na environment Fuel Cladding Chemical Interaction Reprocessing
Stability at high temperature Phase transformation
Neutrons transparency Thermal conductivity
ODS ferritic- Martensitic steels Nano dispersion
ASTRID – 1 er cores 1 5- 1 5 Ti AIM 1
Advanced Austenitic steels, …
EMIR - Orsay - 21 october 2011 Nuclear Materials Department Ph. Dubuisson - 11
Dose 60 dpa Stress few MPa Temperatures 500 – 1100°C 1600°C accident
GFR Cladding – Material Choice
SiC/SiC composite
Key "technological" issues Low deformation Swelling, Irradiation Creep Thermal creep
Elaboration Weldability
Mechanical prop. before, under and after irradiation toughness, embrittlement under irradiation
Behaviour in He environment Fuel Cladding Chemical Interaction Reprocessing
Stability at high temperature Leak-tightness barrier to the fission products
Neutron transparency Thermal conductivity
SiC best candidate despite few drawbacks
3 years
V alloys
EMIR - Orsay - 21 october 2011 Nuclear Materials Department Ph. Dubuisson - 12
Ceramic cladding
GFR Cladding - SiCf/SiC Composites
► Choose the SiC fibers
Hi- Nicalon S , Tyranno SA3, …
Very ceramic sleeving
Gaz (He) tightness Erosion – Oxydation
Dimension
SiCf/SiC Composite Structural properties
ductility yield stress
Separation of the functions Multi- layers
Different layers « porous - dense »
or metallic liner Fission Products tightness
Fuel Compatibility ► Fibrous architecture
2. Multi- layer weaving Mechanical behavior
1 . Filament rolling up
Dimension High density of fibers
low porosity
EMIR - Orsay - 21 october 2011 Nuclear Materials Department Ph. Dubuisson - 13
"New" material for Fast Reactor Components SFR Heat Exchangers GFR Vessel
Interest of FM Steels - 9Cr steels
• Thermal coefficient of conductivity • High thermal dilation coefficient • Good mechanical properties at moderate temperatures • Manufacturing cost
Mechanical properties creep, fatigue, creep- fatigue, fracture mechanic Base metal, HAZ, welded zone predict the long term behavior up to 60 years 540 000 h Weldability homogeneous and heterogeneous Liquid and materials interactions Na, He, … H2O, Vapor
Improvement in toughness
PWR Vessel
0
40
80
120
160
200
1940 1960 1980 2000
Contrainte (Mpa)
provoquant la rupture après
100 000 heures à 600°C
2,25Cr 1Mo9Cr 1Mo (EM10)
12Cr
2,25Cr 1Mo (V)9Cr 2Mo (V,Nb) (EM12)
12Cr 1Mo (V)
2,25Cr 2W (V,Nb)9Cr 1Mo (V,Nb) (T91)12Cr 1Mo 1W (V,Nb)
9Cr 0,5Mo 1/2W (V,Nb)HCM12A: 12Cr 0.5Mo 2W 1Cu (V,Nb)
12Cr 0,5Mo 2W (V,Nb)
12Cr W Co (V,Nb,B)
Diminution des teneurs en MoAugmentation des teneurs en W
Ajout de Mo, V, Nb
Introduction de Co, B
Augmentation des teneurs en W
Optimisation des teneurs en V et Nb
Remplacement partiel du Mo par du W
Historique du développementDes aciers Fe-9/12Cr pour les
Centrales thermiques
0
40
80
120
160
200
1940 1960 1980 2000
Contrainte (MPa)
provoquant la rupture après
600600°°C C -- 100 000 h100 000 h
2,25Cr 1Mo9Cr 1Mo (EM10)
12Cr
2,25Cr 1Mo (V)9Cr 2Mo (V,Nb) (EM12)
12Cr 1Mo (V)
2,25Cr 2W (V,Nb)9Cr 1Mo (V,Nb) (T91)12Cr 1Mo 1W (V,Nb)
9Cr 0,5Mo 1/2W (V,Nb)HCM12A: 12Cr 0.5Mo 2W 1Cu (V,Nb)
12Cr 0,5Mo 2W (V,Nb)
12Cr W Co (V,Nb,B)
Diminution des teneurs en MoAugmentation des teneurs en W
Ajout de Mo, V, Nb
Introduction de Co, B
Augmentation des teneurs en W
Optimisation des teneurs en V et Nb
Remplacement partiel du Mo par du W
Historique du développementDes aciers Fe-9/12Cr pour les
Centrales thermiques
0
40
80
120
160
200
1940 1960 1980 2000
Contrainte (Mpa)
provoquant la rupture après
100 000 heures à 600°C
2,25Cr 1Mo9Cr 1Mo (EM10)
12Cr
2,25Cr 1Mo (V)9Cr 2Mo (V,Nb) (EM12)
12Cr 1Mo (V)
2,25Cr 2W (V,Nb)9Cr 1Mo (V,Nb) (T91)12Cr 1Mo 1W (V,Nb)
9Cr 0,5Mo 1/2W (V,Nb)HCM12A: 12Cr 0.5Mo 2W 1Cu (V,Nb)
12Cr 0,5Mo 2W (V,Nb)
12Cr W Co (V,Nb,B)
Diminution des teneurs en MoAugmentation des teneurs en W
Ajout de Mo, V, Nb
Introduction de Co, B
Augmentation des teneurs en W
Optimisation des teneurs en V et Nb
Remplacement partiel du Mo par du W
Historique du développementDes aciers Fe-9/12Cr pour les
Centrales thermiques
0
40
80
120
160
200
1940 1960 1980 2000
Contrainte (MPa)
provoquant la rupture après
600600°°C C -- 100 000 h100 000 h
2,25Cr 1Mo9Cr 1Mo (EM10)
12Cr
2,25Cr 1Mo (V)9Cr 2Mo (V,Nb) (EM12)
12Cr 1Mo (V)
2,25Cr 2W (V,Nb)9Cr 1Mo (V,Nb) (T91)12Cr 1Mo 1W (V,Nb)
9Cr 0,5Mo 1/2W (V,Nb)HCM12A: 12Cr 0.5Mo 2W 1Cu (V,Nb)
12Cr 0,5Mo 2W (V,Nb)
12Cr W Co (V,Nb,B)
Diminution des teneurs en MoAugmentation des teneurs en W
Ajout de Mo, V, Nb
Introduction de Co, B
Augmentation des teneurs en W
Optimisation des teneurs en V et Nb
Remplacement partiel du Mo par du W
Historique du développementDes aciers Fe-9/12Cr pour les
Centrales thermiques
Evolution of 9-12 Cr martensitic steels
600°C 100 000 h
Stress to creep rupture
EMIR - Orsay - 21 october 2011 Nuclear Materials Department Ph. Dubuisson - 14
SFR – Main components
Core Sub-assemblies
400 – 650° C Irradiation
Life time to design
30 60 years
5 105 h
Vessel 400°C
No deformation Negligeable creep
Bottom core structures Int. Heat Exchangers, Pumps
Cold structures 400°C No deformation low irradiation
Upper core structures Hot structures 550°C
Creep, Weld joint behavior low irradiation
Steam Generators, Heat Exchangers 350 – 525° C
Aging, Welds, Compatibility Avoid Na – H2O
Circuits - Pipes 350 – 550° C
Creep, fatigue, creep-fatigue,
thermal fatigue,… Aging Welds
9 Cr F/M ODS Adv. Aust.
31 6 LN 31 6 LN
31 6 LN 800H
Ni alloys
9 Cr 800H 31 6 LN
9 Cr 31 6 LN
1000 MWe Pool type
Modular SG AREVA design
EMIR - Orsay - 21 october 2011 Nuclear Materials Department Ph. Dubuisson - 15
a
f
g
h
b
c
d
e
1
2
3
4 5
6
7
8 9
10 11
TF coils
central ports
Vacuum vessel
Blanket module
divertor plates
shield lower ports
upper ports
blanket manifolds
Materials for Fusion
W & CC tiles
ITER 316LN/CuCrZr/Be TBM Eurofer DEMO,… ODS, SiC/SiC, V alloys
10-20 appm He/dpa ~ 45 appm H/dpa
Eurofer SiC/SiC
ODS layers
316 LN < 4 dpa 650°C Eurofer 3 - 80 dpa 550°C Eurofer ODS 3 - 80 dpa 650°C Ferritic ODS 200 dpa 800°C V alloys high dose 700°C W alloys low dose > 1000°C SiC/SiC high dose 1000°C
50 – 80 dpa 100-150 dpa
Tmax
EMIR - Orsay - 21 october 2011 Nuclear Materials Department Ph. Dubuisson - 16
Accelerator I ~ 10 - 30 mA E ~ 1 GeV
Fluid Pb-Bi, He,
Na
Window 9% Cr martensitic steels 250 – 550°C ?
few 10 dpa /year
Materials for Accelerator Driven System
100
101
102
103
104
105
H HeLi Be B C N O F Ne NaMg Al Si P S Cl Ar K Ca Sc Ti V CrMnFe
appm
Protons 1 GeV, 58 µA/cm2, 200 Jepp
H effect
Hardening Embrittlement ?
Intergranular embrittlement ?
He embrittlement
window
EMIR - Orsay - 21 october 2011 Nuclear Materials Department Ph. Dubuisson - 17
Our path forward How to develop, optimize and qualify
in timely fashion the materials required ?
Multiscale modelling
Experimental simulation Charged particles – multi beams ions, electrons JANNUS, GANIL, HVEM,…
"Smart" experiments in MTR,… Osiris, HFR, … RJH, IFMIF, … Astrid, Allegro,…
Shorten the development time of new materials
Predict the behavior of materials under conditions not or hardly accessible to experiments (long times,...)
Space
1 nm 1µm
Time
1 s
1 ps40 nm
Atoms
Electrons
Microstructure
10 mm
Dislocations andirradiation defects
Structure
1mm
1 y
1m
1 c
1 mm
10 m
Electronic StructureMolecular Dynamics
Object or Event Monte Carlo
Crystal Plasticity (CP)Homogenization
Finite Elements (EF)
Monte CarloClusters Dynamics
Formation and mobilityof point defects (dp)
Evolution of Dislocations Networkdefects clusters, solute
Dislocations Dynamics (DD)
TEM
TAP
SANS
MechanicalTests
SEMEBSD
Déformation
Behavior rulesMécanique de la rupture
Modeling
Ab initio
EMIR - Orsay - 21 october 2011 Nuclear Materials Department Ph. Dubuisson - 18
Materials for Reactor Systems
France France France France
GEN IV
PWR
Water 290 – 330°C
155 bars
Clad Zr Alloys
10 - 15 dpa 0.1 dpa 120 dpa
France France
Vessel 16MND5
Internals 304 – 316
300 – 400
H & He production
Current fleet PWR
1975 2000 2025 2050 2075
Gen IV
EPR
Life time extension
Reactors
EMIR - Orsay - 21 october 2011 Nuclear Materials Department Ph. Dubuisson - 19
Space
1 nm 1µm
Time
1 s
1 ps40 nm
Atoms
Electrons
Microstructure
10 mm
Dislocations andirradiation defects
Structure
1mm
1 y
1m
1 c
1 mm
10 m
Electronic Structure Molecular Dynamics
Object or Event Monte Carlo
Crystal Plasticity (CP) Homogenization
Finite Elements (EF)
Monte Carlo Clusters Dynamics
Formation and mobility of point defects (dp)
Evolution of Dislocations Network defects clusters, solute
Dislocations Dynamics (DD)
TEM
TAP
SANS
Irradiations by charged particles
JANNUS, GANIL, electrons, …
Experimental reactors MTR – Osiris, RJH,HFIR,…
FR – Phénix, BN 600, ASTRID, Allegro, …
Mechanical Tests
SEM EBSD
Déformation
Behavior rules Mécanique de la rupture
Charaterizations At same scale
Irradiations
Modeling
Ab initio
Multi-scale Modeling in Irradiation Effects
EMIR - Orsay - 21 october 2011 Nuclear Materials Department Ph. Dubuisson - 20
Tools Multi-scale modelling
Ab initio, Molecular Dynamic, Rate theory • Prediction of the irradiation effects on materials • Orientation of experience and characterization
steels, ceramics, composites, fuel Simulation by charged particles ions or electrons
Fundamental mechanisms and physical modeling
Tests in experimental reactors Phénix Osiris RJH ASTRID, Allegro, …
BOR 60, BN 600, Monju, HFIR,…
Single Beam Irradiation
Triple beam irradiation
Ion Beam Analysis
Épiméthée3 MV
PelletronECR Source
Yvette 2.5 MV Van de Graaff
Japet 2 MV Tandem
JANNUS - ions Triple beam irradiation
HVEM – electrons 1.2 MeV