Tensile properties and microstructure of austenitic steels irradiated in different reactors

1. Tensile properties and microstructure of austenitic steels irradiated in different reactors Ph. DubuissonX. Averty V.K. ShamardinV.I. Prokhorov J.P. MassoudC. Pokor M. Žamboch Y. Bréchet Influence of Atomic Displacement Rate onRadiation-induced Aging of Power Reactor Ulianovsk, Russia - October 2 - 8, 2005

2. Objective : Evaluate the effects of neutron irradiation on mechanical properties and resistance to SCC EBR-II/Phénix Bor-60BORIS Irradiations in Experimental Reactors Reach "rapidly" the end-of-life doses FBR • Mechanical properties • SCC • Microstructure • Modelling Temperature 370°C Tensile specimens 360°C Materials Representative of Core Internals of the PWRs • SA 304L Baffle plates, Former, Core barrel • CW 316 Baffle bolts PWR40 years 320°C 300°C Dose » 30 dpa » 95 dpa

3. EBR-II/Phénix 2 irradiation areas Osiris Area without shield Fast & thermal neutrons Bor-60 BORIS Tensile SM Area with Hf shield Fast neutrons Steel Irradiations in Experimental Reactors Irradiations in FBR  Spectrum effect ? Flux, Gaz He, H Temperature 370°C 360°C PWR40 years • BOR-60 / Osiris Tensile 10 dpa 320°C • He effect • Mechanical properties • SCC tests 300°C SAMARA Dose • Modelling » 30 dpa » 95 dpa

4. CW 316 SA 304L Tensile propertiesFast Breeder Reactor BOR 60 320°C Te = 330°C 3 10-4 s-1 SA 304L BOR 60 CW 316 Total Elongation5 – 10% U.T.S.  YS0.2% SA 304L > 5 dpaCW 316  10 dpa Saturation CW 316 > SA 304L - 125 dpa

5. Osiris Osiris BOR 60 BOR 60 Tensile propertiesBOR 60 - Osiris Te = 330°C 3 10-4 s-1 5 dpa 10 dpa SA 304L No difference between Osiris and BOR 60 No effect of neutron spectrum Saturation of mechanical properties > 5 dpa

6. Tensile propertiesBOR 60 - Osiris 320°C Te = 330°C 3 10-4 s-1 CW 316 SA 304L BOR 60 SA 304L Osiris CW 316 No difference between Osiris and BOR 60No effect of neutron spectrum Saturation of mechanical properties SA 304L 5 dpa CW 316 10 dpa

7. Tensile propertiesHelium effect SM 2 300°C Te = 330°C 3 10-4 s-1 Same Flux 6 dpa 17 dpa SA 304L 600 appm He 10 appm He 14 appm He 300 appm He SM 2 No obviouseffect of Helium (H2) content Saturation of mechanical properties < 6 dpa

8. Tensile propertiesBOR 60 – Osiris – SM 2 No obvious effect of helium (H2) on tensile characteristics Tensile characteristics similar to those measured after irradiationin Bor-60 (FBR) at 320°C both for CW 316 and SA 304L Te = 330°C 3 10-4 s-1 CW 316 BOR 60 SA 304L SA 304L Osiris SM 2 CW 316 No flux effect

9. Tensile properties • Saturation dose at 5 dpa for SA 304L10 dpa for CW 316 • No evolution between 10 and 125 dpa for both SA 304L - CW 316 • CW 316 > SA 304L hardness – residual ductility • No neutron spectrum effect on tensile characterictics Gaz content and flux

10. PWR Internals Hardening Model Evolution of the Yield Stress after irradiation Temperature, fluence, neutron spectrum Model of the population of point defects clusters (dislocation loops) Microstructural data ofneutron irradiated materials TEM Model of hardening by a cluster population Dsproportionala, L Yield Strength of neutron irradiated materials Tensile tests EBR II, Osiris, BOR 60

11. EBR II375°C - 10 dpa Voids CW 316 SA 304L SA 304L 375°C 330°C 330°C 375°C Microstructure Frank Loops Black dots No precipitationNo more dislocation lines EBR IIOsiris BOR 60 Main featureFrank loops formation Saturation for dose about 5 - 10 dpa size SA 304L  316 density SA 304L > CW 316

12. Interstitial or vacancy Interstitial or . vacancy Interstitial vacancycluster sinks : clusters dislocation lines grain boundaries / free surfaces Evolution of the concentration of interstitial or vacancy cluster containing n defects Microstructure ModellingFrank loop Loops Chemical kinetic Model"Cluster Dynamics" External source of irradiation defects Evolution of the concentration of point defects Neutron spectrum Flux - EPKA • Production • - Recombination (v-i) • Loss of v and i at sinks • Agglomeration Neutrons Dislocation network evolution Homogeneous medium

13. Interstitial or vacancy Interstitial or . vacancy Interstitial vacancycluster sinks : clusters dislocations grain boundaries / free surfaces Microstructure ModellingFrank loop Loops Chemical kinetic Model"Cluster Dynamics" External source of irradiation defects Neutrons Homogeneous medium

14. SA 304L 375°C 330°C Microstructure ModellingFrank loops Chemical kinetic Model"Cluster Dynamic" Emv 1.35 eV Emi 0.45 eV Eb2i 0.6 eV r0 1010 cm-2 • Adjust material parameters of the model on low dose data EBR II - Osiris • Predict the behavior at higher doses EBR II – Osiris – BOR 60 • Comparison with experimental data – BOR 60 • Comparison with future results high doses BOR 60 (90 dpa) and Osiris (10 dpa)

15. Experimental Simulation Temperature dose Material (°C) (dpa) - 3 - 3 - 3 f r f r f r (nm) (m ) (nm) (m ) (nm) (m ) i i v v i i 21 21 320 10 11,5 25 10 / / 11,2 27 10 2 1 20 21 CW 316 320 19 7 92 10 2 < 10 12,8 32 10 21 333 24 10 12 10 / Microstructure of expertised components 21 / 18,7 18 10 21 21 21 SA 304 310 35 10 15 10 1 1 10 10,2 80 10 In agreement with results from experimental reactors In terms of interstitial loops size and density, the results of the model are in relatively good agreementwith the results obtained from field experience in PWRs. IV – Back to field experience Chap IV-1

16. Ds SA 304L 330°C = a r f Ds + M mb loops loops loops BOR 60 Osiris al ( ) ? Model of defect clusters no diameter and density saturation Orowan hardening no hardening saturation Hardening - Orowan Model Cluster DynamicsModel Dislocations networkevolution Good agreement at low dose Data / Model aloops 0,4 Same for CW 316

17. Ds No Defaulting Defaulting dose Modified Orowan Model • Hardening due to Frank Loops • Defaulting of the Frank loops • Transformation in perfect loop under a applied stress Critical shear stress for defaulting a Frank loop of diameter fOne relation between r and f • r and f increase with dose • Perfect loop glide and annihilate • Saturation at the critical stress • Critical dose for the mechanism of hardening dc Main parameters g : Stacking Fault Energy, rd One adjustable parameterNumber of dislocations in the pile-up

18. Ds Ds CW 316 330°C 330°C 375°C 375°C SA 304L Modified "Orowan" model Hardening model permitting defaulting of Frank Loops  Saturation of Hardening Voids r0 1014 cm-2g 42 Jm-2 r0 1010 cm-2g 26 Jm-2 330°C Good description of experimental data 375°C Need data at high dose to verify Voids  data / Model in SA 304  2 steels : r0g Well description of experimental data by the model

19. 13 MPa O2 ~ 0 ppb H2 29 – 30 ml/kg <10 ppb Slow Strain Rate Tests (SSRT)PWR environment Flow rate2 autoclave vol./h Te = 320°C 5 10-8 s-1 T.E. of SSRT specimens strongly reduced (compared to tensile tests in air) lower for the specimens with “low helium” Hardening lower for SA 304 after tests in PWR No significant difference in susceptibility between SA304L and CW 316 SA 304L s MPa CW 316 s MPa 300°C Air PWR Air 5 dpa PWR Slight effectof He content SA 304L CW 316

20. TG / IGfracture ductile fracture ductile fracture IGfracture Fracture surfaces 5 facets SA 304L CW 316 3 facets He (appm) 294 9 297 15 % brittle fracture 33,3 71,1 38,0 50,8 Low He facet initiation TG IG TG»IG IG fracture in facets TG»IG IG>TG TG=IG IG>TG high He Transgranular Intergranular

21. Conclusions - Perspectives No evolution between10 and 125 dpa • Tensile properties • Saturation dose at 5 dpa for SA 304L10 dpa for CW 316 • CW 316 > SA 304L hardness – residual ductility • No neutron spectrum effect on tensile characterictics Gaz content and flux • Microstructure • High density of small Frank loops + Voids at high temperature in SA 304L • Disappearance of the initial dislocations network • No precipitation • Reproduce Microstructure observed on PWR components • Hardening Model • Cluster Dynamic Model Good agreement with TEM quantification – Frank loopsNo real saturation of loop number density and diameter • Hardening Model • Orowan Model No saturation of hardening • Modified Orowan Model permitting the defaulting of Frank loopsSaturation of Hardening

22. Conclusions - Perspectives • In simulated PWR water • Total Elongation of the SSRT specimens strongly reduced • Fracture surface partly intergranular • T.E. lower - Fracture surfaces more intergranular “low helium” content • Further examinations and SCC tests will be performed on more highly irradiated materials • Mechanical properties saturate • He content increases • Intergranular fracture SM 2 > BOR 60 Flux effect ? Medium ?

23. This work was performed through a collaboration betweenEDF, CEA and RIAR partly sponsored by EPRI Authors are grateful to: HT Tang (EPRI), V. Golovanov and G. Shimansky (RIAR), P. Brabec and A. Brožova (NRI) F. Rozenblum, J.C. Brachet and A. Barbu (CEA).