challenge of materials for nuclear reactors -...

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


Outline

Requirements for nuclear materials non fissile Gen II-III

Gen IV

Fusion – Accelerator Driven System

Conclusions


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


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


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


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


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


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


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


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, …


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


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


"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)


600600°°C C -- 100 000 h100 000 h


12Cr


12Cr 1Mo (V)




12Cr W Co (V,Nb,B)


Ajout de Mo, V, Nb







0

40

80

120

160

200

1940 1960 1980 2000

Contrainte (Mpa)


100 000 heures à 600°C


12Cr


12Cr 1Mo (V)




12Cr W Co (V,Nb,B)


Ajout de Mo, V, Nb







0

40

80

120

160

200

1940 1960 1980 2000

Contrainte (MPa)


600600°°C C -- 100 000 h100 000 h


12Cr


12Cr 1Mo (V)




12Cr W Co (V,Nb,B)


Ajout de Mo, V, Nb







Evolution of 9-12 Cr martensitic steels

600°C 100 000 h

Stress to creep rupture


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


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


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


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


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


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


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

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