Challenge of materials for nuclear reactors fission and fusion

Transcription

1 EMIR Users days october Challenge of materials for nuclear reactors fission and fusion Ph. Dubuisson, P. Yvon Nuclear Materials Department Orsay France 21 october Ph. Dubuisson - 1

2 Outline Requirements for nuclear materials non fissile Gen II-III Gen IV Fusion Accelerator Driven System Conclusions Ph. Dubuisson - 2

3 Requirements for materials years life time Vessel, internals, Fast neutron damage Evolution of the microstructure phase instability, precipitation, voids, amorphization, Dimensionnal changes Modification of mechanical properties swelling, growth, irradiation & thermal creep YS, UTS, elongation, toughness, hardening, embrittlement, Thermal creep time to rupture hydrogen & helium Embrittlement Resistance at high temperature Mechanical properties Embrittlement at high temperature Thermal creep Creep - fatigue interaction YS, UTS, elongation, toughness, time to rupture, deformation Compatibility with the different environments primary and secondary fluid, fuel, reprocessing, Fuel Cladding Chemical interaction, Hydrogen cracking Corrosion & cracking by stress corrosion point defects & clusters gas, transmutation also incidental and accidental conditions heat Exchangers I- SCC, IASCC, Ph. Dubuisson - 3

4 Requirements for materials Complementary considerations Availability and cost of materials Fabricability, joining technology Low activity welding, 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 Ph. Dubuisson - 4

5 GEN 2&3 - PWR - Irradiated Components Extension of lifetime Core Internals Nickel alloys C 10/15 dpa 5 6 years ~ 300 C 0.1 dpa years Fuel Assemblies Zr alloys 155 bars 293 C Water H 2, LiOH, B Vessel Bainitic steel 1 6MND5 A508 Cl 3 neutrons temperature mechanical stresses environment 328 C ~ 320 C few 0.1 dpa years Control rods Austenitic steels loops ~ 320 C ~ 10 dpa few years Core Internals Austenitic steels C dpa years Ph. Dubuisson - 5

6 PWR - Vessel steels Toughness MPa m Start of life PWR Vessel bainitic steel 16 MND5 (A508 CI.3) Irradiation DBTT Shift Decrease in USE Hardening Loss in ductility TT TT Temperature C TT ageing irradiation 90 nm GPM Rouen VVER steels Cu P Mn Ni Si Clusters + point defect clusters Fluence en fin de vie des unités 900 MWe (40 ans) 19 2 Fluence 10 n/cm (E>1MeV) Ph. Dubuisson - 6

7 PWR cladding Zr alloys Contrainte (MPa) σ MPa IPG Servic e EtLD Service storage PCI RIA RIA DENO Transport Dry-out shipment APRP High density of <a> loop Localization of deformation Température ( C) LOCA,T ε,t ε,t ε,t ε very low low high very high T C life time X2 in 15 years Ph. Dubuisson - 7

8 GEN IV 6 systems Irradiation conditions New goals for sustainable nuclear energy New challenges for materials! Here normal operating conditions Température ( C) Supercritical Water cooled Reactor T ( C) VHTR Générations II- III Gas Fast Reactor Molten Salt Reactor dose (dpa) Sodium Fast Déplacement par atome (dpa) Lead Fast Reactor Reactor The most mature option Ph. Dubuisson - 8

9 SFR Cladding Material Choice Dose > 150 dpa Stress 100 MPa Temperatures C 8 10 years Improved safety - Reduce Fuel enrichment Reactivity excess Potential void effect Low deformation Swelling, Irradiation Creep Thermal creep Maximise fuel content reduce Na in the core SFR SPX (%) Average 316 Ti Average 15/15Ti Best lot of 15/15Ti Phénix ε (%) Fe-18Cr 650 C 1 80 MPa PHENIX 4 2 Ferritic-martensitic (F/M) 2 steels, ODS included 0 dose (dpa) ODS Fe-18Cr 0,5 Y 2 O 3 Time (h) 0,0001 0,001 0,01 0, ODS ferritic- Martensitic steels Nano dispersion Ph. Dubuisson - 9

10 SFR Cladding Material Choice Dose > 150 dpa Stress 100 MPa Temperatures C 8 10 years Neutrons transparency Thermal conductivity ASTRID 1 er cores Ti AIM 1 Advanced Austenitic steels, Improved safety - Reduce Fuel enrichment Reactivity excess Potential void effect Key "technological" issues Low deformation Swelling, Irradiation Creep Thermal creep Elaboration Weldability ODS ODS ferritic- Martensitic Martensitic 9Cr steels steels ODS Ferritic Nano dispersion Cr steels Nano dispersion Maximise fuel content reduce Na in the core 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 SFR SPX1 PHENIX Ph. Dubuisson - 10

11 GFR Cladding Material Choice Dose 60 dpa Stress few MPa Temperatures C 1600 C accident Neutron transparency Thermal conductivity Key "technological" issues Low deformation Swelling, Irradiation Creep Thermal creep Elaboration Weldability 3 years 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 SiC best candidate despite few drawbacks SiC/SiC composite V alloys Ph. Dubuisson - 11

12 GFR Cladding - SiC f /SiC Composites Separation of the functions Multi- layers Different layers «porous - dense» or metallic liner Fission Products tightness Fuel Compatibility Choose the SiC fibers Hi- Nicalon S, Tyranno SA3, Fibrous architecture 1. Filament rolling up SiC f /SiC Composite Structural properties ductility yield stress Very ceramic sleeving Dimension High density of fibers low porosity 2. Multi- layer weaving Mechanical behavior Ceramic cladding Gaz (He) tightness Erosion Oxydation Dimension Ph. Dubuisson - 12

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 PWR Vessel Improvement in toughness Mechanical properties creep, fatigue, creep- fatigue, fracture mechanic Base metal, HAZ, welded zone predict the long term behavior up to 60 years h Weldability homogeneous and heterogeneous Liquid and materials interactions Na, He, H 2 O, Vapor Stress to creep rupture Contrainte (Mpa) (MPa) provoquant la rupture après C heures 000 h à 600 C Optimisation des teneurs en V et Nb Remplacement partiel du Mo par du W 2,25Cr 1Mo 9Cr 1Mo (EM10) 12Cr 600 C h 2,25Cr 1Mo (V) 9Cr 2Mo (V,Nb) (EM12) 12Cr 1Mo (V) Ajout de Mo, V, Nb Introduction de Co, B Augmentation des teneurs en W 9Cr 0,5Mo 1/2W (V,Nb) HCM12A: 12Cr 0.5Mo 2W 1Cu (V,Nb) 12Cr 0,5Mo 2W (V,Nb) 2,25Cr 2W (V,Nb) 9Cr 1Mo (V,Nb) (T91) 12Cr 1Mo 1W (V,Nb) 12Cr W Co (V,Nb,B) Diminution des teneurs en Mo Augmentation des teneurs en W Evolution of 9-12 Cr martensitic steels Historique du développement Des aciers Fe-9/12Cr pour les Centrales thermiques Ph. Dubuisson - 13

14 1000 MWe Pool type Modular SG AREVA design 9 Cr 31 6 LN Circuits - Pipes C Creep, fatigue, creep-fatigue, thermal fatigue, Aging Welds SFR Main components Steam Generators, Heat Exchangers C Aging, Welds, Compatibility 9 Cr 800H 31 6 LN Avoid Na H 2 O 31 6 LN 800H Ni alloys Upper core structures Hot structures 550 C Creep, Weld joint behavior low irradiation Life time to design years 9 Cr F/M ODS Adv. Aust h Core Sub-assemblies C Irradiation Bottom core structures Int. Heat Exchangers, Pumps Cold structures 400 C No deformation low irradiation Vessel 400 C No deformation Negligeable creep 31 6 LN 31 6 LN Ph. Dubuisson - 14

15 Materials for Fusion Vacuum vessel blanket manifolds b c a d e TF coils f g upper ports Blanket module central ports ODS layers Eurofer SiC/SiC dpa dpa shield ITER TBM DEMO, divertor plates W & CC tiles h lower ports 316LN/CuCrZr/Be Eurofer ODS, SiC/SiC, V alloys T max 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 appm He/dpa ~ 45 appm H/dpa Ph. Dubuisson - 15

16 Materials for Accelerator Driven System Accelerator I ~ ma E ~ 1 GeV Window 9% Cr martensitic steels C? H effect He embrittlement appm few 10 dpa /year Intergranular embrittlement? Hardening Embrittlement? 10 3 Fluid Pb-Bi, He, Na H He Li Be B C N O F Ne NaMg Al Si P S Cl Ar K Ca Sc Ti V CrMnFe window Protons 1 GeV, 58 µa/cm 2, 200 Jepp Ph. Dubuisson - 16

17 Our path forward How to develop, optimize and qualify in timely fashion the materials required? 1 c 1 y 1 s 1 ps Time TAP Electrons SANS Atoms 40 nm TEM Dislocations and irradiation defects Electronic Structure Ab initio 1 nm 1µm 10 mm SEM EBSD Microstructure Formation and mobility of point defects (dp) Mechanical Tests Molecular Dynamics Object or Event Monte Carlo 1mm Structure Déformation 1 mm Evolution of Dislocations Network defects clusters, solute 1m Space Multiscale modelling Modeling Behavior rules 10 m Mécanique de la rupture Finite Elements (EF) Crystal Plasticity (CP) Homogenization Dislocations Dynamics (DD) Monte Carlo Clusters Dynamics Shorten the development time of new materials Predict the behavior of materials under conditions not or hardly accessible to experiments (long times,...) "Smart" experiments in MTR, Osiris, HFR, RJH, IFMIF, Astrid, Allegro, Experimental simulation Charged particles multi beams ions, electrons JANNUS, GANIL, HVEM, Ph. Dubuisson - 17

18 Materials for Reactor Systems GEN IV H & He production PWR Water C 155 bars Clad Zr Alloys Vessel 16MND5 Internals dpa 0.1 dpa 120 dpa Reactors Current fleet PWR Life time extension EPR Gen IV France France France France France France Ph. Dubuisson - 18

19 Multi-scale Modeling in Irradiation Effects 1 c 1 y Irradiations by charged particles JANNUS, GANIL, electrons, Time SANS Experimental reactors MTR Osiris, RJH,HFIR, FR Phénix, BN 600, ASTRID, Allegro, TEM Dislocations and irradiation defects SEM EBSD Microstructure Mechanical Tests Structure 10 m Irradiations Charaterizations At same scale Modeling Behavior rules Mécanique de la rupture Finite Elements (EF) 1 s TAP Atoms 1 mm Déformation Crystal Plasticity (CP) Homogenization 1 ps Electrons 40 nm Electronic Structure Ab initio 1 nm 1µm 10 mm Evolution of Dislocations Network defects clusters, solute Formation and mobility of point defects (dp) Molecular Dynamics Object or Event Monte Carlo 1mm Space Ph. Dubuisson m Dislocations Dynamics (DD) Monte Carlo Clusters Dynamics

20 Multi-scale modelling Tools 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 Japet 2 MV Tandem Triple beam irradiation Single Beam Irradiation HVEM electrons 1.2 MeV Tests in experimental reactors Phénix Osiris RJH ASTRID, Allegro, JANNUS - ions Triple beam irradiation Ion Beam Analysis Yvette 2.5 MV Van de Graaff Épiméthée 3 MV Pelletron ECR Source BOR 60, BN 600, Monju, HFIR, Ph. Dubuisson - 20