DEN - Saclay – Nuclear Material departement "Influence of atomic displacement rate… " Workshop - Ulyanovsk - October 5, 2005 Ph. Dubuisson - 1 DMN / SRMA Tensile properties and microstructure of austenitic steels irradiated in different reactors Ph. Dubuisson X. Averty M. Žamboch V.K. Shamardin V.I. Prokhorov J.P. Massoud C. Pokor Influence of Atomic Displacement Rate on Radiation-induced Aging of Power Reactor Ulianovsk, Russia - October 2 - 8, 2005 Y. Bréchet

DEN - Saclay – Nuclear Material departement "Influence of atomic displacement rate… " Workshop - Ulyanovsk - October 5, 2005 Ph. Dubuisson - 2 DMN / SRMA FBR Reach "rapidly" the end-of-life doses  FBR Mechanical properties SCC Microstructure Modelling Objective Objective : Evaluate the effects of neutron irradiation on mechanical properties and resistance to SCC Irradiations in Experimental Reactors Temperature 370°C 300°C 320°C BORIS Bor-60 BORIS EBR-II/Phénix  30 dpa Dose PWR 40 years  95 dpa 360°C Materials Representative of Core Internals of the PWRs  SA 304LBaffle plates, Former, Core barrel  CW 316Baffle bolts CCrNiMnMoSiCu SA 304L CW Tensile specimens

DEN - Saclay – Nuclear Material departement "Influence of atomic displacement rate… " Workshop - Ulyanovsk - October 5, 2005 Ph. Dubuisson - 3 DMN / SRMA Irradiations in Experimental Reactors Irradiations in FBR  Spectrum effect ?  He effect Mechanical properties SCC tests  BOR-60 / Osiris Tensile 10 dpa Area without shield Fast & thermal neutrons 2 irradiation areas Tensile Area with Hf shield Fast neutrons Steel Temperature 370°C 300°C 320°C BORIS Bor-60 BORIS EBR-II/Phénix  30 dpa Dose PWR 40 years  95 dpa 360°C SM Osiris Flux, Gaz He, H He, H  Modelling SAMARA

DEN - Saclay – Nuclear Material departement "Influence of atomic displacement rate… " Workshop - Ulyanovsk - October 5, 2005 Ph. Dubuisson - 4 DMN / SRMA Tensile properties Fast Breeder Reactor BOR 60 T e = 330°C s °C BOR 60 SA 304L CW 316 SA 304L CW 316 Saturation Total Elongation 5 – 10% U.T.S.  YS 0.2% CW 316 > SA 304L dpa SA 304L > 5 dpa CW 316  10 dpa

DEN - Saclay – Nuclear Material departement "Influence of atomic displacement rate… " Workshop - Ulyanovsk - October 5, 2005 Ph. Dubuisson - 5 DMN / SRMA BOR 60 Tensile properties BOR 60 - Osiris Osiris 5 dpa Osiris 10 dpa No difference between Osiris and BOR 60 No effect of neutron spectrum Saturation of mechanical properties > 5 dpa T e = 330°C s -1 SA 304L

DEN - Saclay – Nuclear Material departement "Influence of atomic displacement rate… " Workshop - Ulyanovsk - October 5, 2005 Ph. Dubuisson - 6 DMN / SRMA Tensile properties BOR 60 - Osiris BOR 60 SA 304L CW 316 SA 304L CW 316 Osiris No difference between Osiris and BOR 60 No effect of neutron spectrum Saturation of mechanical properties SA 304L5 dpa CW dpa T e = 330°C s °C

DEN - Saclay – Nuclear Material departement "Influence of atomic displacement rate… " Workshop - Ulyanovsk - October 5, 2005 Ph. Dubuisson - 7 DMN / SRMA Tensile properties Helium effect SM 2 SM 2 6 dpa 17 dpa Same Flux 10 appm He 300 appm He No obvious effect of Helium (H 2 ) content Saturation of mechanical properties < 6 dpa T e = 330°C s °C 14 appm He 600 appm He SA 304L

DEN - Saclay – Nuclear Material departement "Influence of atomic displacement rate… " Workshop - Ulyanovsk - October 5, 2005 Ph. Dubuisson - 8 DMN / SRMA Tensile properties BOR 60 – Osiris – SM 2 BOR 60 SA 304L CW 316 SA 304L CW 316 Osiris T e = 330°C s -1 No obvious effect of helium (H 2 ) on tensile characteristics Tensile characteristics similar to those measured after irradiation in Bor-60 (FBR) at 320°C both for CW 316 and SA 304L SM 2 No flux effect

DEN - Saclay – Nuclear Material departement "Influence of atomic displacement rate… " Workshop - Ulyanovsk - October 5, 2005 Ph. Dubuisson - 9 DMN / SRMA Tensile properties  Saturation dose at 5 dpa for SA 304L 10 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

DEN - Saclay – Nuclear Material departement "Influence of atomic displacement rate… " Workshop - Ulyanovsk - October 5, 2005 Ph. Dubuisson - 10 DMN / SRMA Hardening Model Evolution of the Yield Stress after irradiation Temperature, fluence, neutron spectrum Model of the population of point defects clusters (dislocation loops) Model of hardening by a cluster population  proportional , L Microstructural data of neutron irradiated materials TEM Yield Strength of neutron irradiated materials Tensile tests PWR Internals EBR II, Osiris, BOR 60

DEN - Saclay – Nuclear Material departement "Influence of atomic displacement rate… " Workshop - Ulyanovsk - October 5, 2005 Ph. Dubuisson - 11 DMN / SRMA Microstructure Frank Loops No precipitation No more dislocation lines Saturation for dose about dpa sizeSA 304L  316 densitySA 304L > CW °C 330°C 375°C 330°C SA 304L Main feature Frank loops formation Black dots Voids SA 304L CW 316 EBR II 375°C - 10 dpa EBR II Osiris BOR 60

DEN - Saclay – Nuclear Material departement "Influence of atomic displacement rate… " Workshop - Ulyanovsk - October 5, 2005 Ph. Dubuisson - 12 DMN / SRMA Microstructure Modelling Frank loop Chemical kinetic Model "Cluster Dynamics" Dislocation network evolution Loops Evolution of the concentration of point defects Production - Recombination (v-i) - Loss of v and i at sinks - Agglomeration Evolution of the concentration of interstitial or vacancy cluster containing n defects External source of irradiation defects Interstitial or. vacancy Interstitial vacancy cluster Interstitial or vacancy sinks : clusters dislocation lines grain boundaries / free surfaces Neutrons Homogeneous medium Neutron spectrum Flux - E PKA

DEN - Saclay – Nuclear Material departement "Influence of atomic displacement rate… " Workshop - Ulyanovsk - October 5, 2005 Ph. Dubuisson - 13 DMN / SRMA Microstructure Modelling Frank loop Chemical kinetic Model "Cluster Dynamics" External source of irradiation defects Interstitial or. vacancy Interstitial vacancy cluster Interstitial or vacancy sinks : clusters dislocations grain boundaries / free surfaces Neutrons Homogeneous medium Loops

DEN - Saclay – Nuclear Material departement "Influence of atomic displacement rate… " Workshop - Ulyanovsk - October 5, 2005 Ph. Dubuisson - 14 DMN / SRMA Microstructure Modelling Frank loops 1.Adjust material parameters of the model on low dose data EBR II - Osiris 2.Predict the behavior at higher doses EBR II – Osiris – BOR 60 3.Comparison with experimental data – BOR 60 4.Comparison with future results high doses BOR 60 (90 dpa) and Osiris (10 dpa) Chemical kinetic Model "Cluster Dynamic" SA 304L 375°C 330°C E m v 1.35 eV E m i 0.45 eV E b 2i 0.6 eV  cm -2

DEN - Saclay – Nuclear Material departement "Influence of atomic displacement rate… " Workshop - Ulyanovsk - October 5, 2005 Ph. Dubuisson - 15 DMN / SRMA IV – Back to field experience / 18, SA , In terms of interstitial loops size and density, the results of the model are in relatively good agreement with the results obtained from field experience in PWRs. Chap IV-1 Microstructure of expertised components In agreement with results from experimental reactors

DEN - Saclay – Nuclear Material departement "Influence of atomic displacement rate… " Workshop - Ulyanovsk - October 5, 2005 Ph. Dubuisson - 16 DMN / SRMA Hardening - Orowan Model Dislocations network evolution Good agreement at low dose Data / Model Cluster Dynamics Model Model of defect clustersno diameter and density saturation Orowan hardeningno hardening saturation  loops  0,4 loops bb   M  +  l ( ) ? Same for CW 316 SA 304L 330°C  Osiris BOR 60

DEN - Saclay – Nuclear Material departement "Influence of atomic displacement rate… " Workshop - Ulyanovsk - October 5, 2005 Ph. Dubuisson - 17 DMN / SRMA  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  One relation between  and    and  increase with dose  Perfect loop glide and annihilate  Saturation at the critical stress  Critical dose for the mechanism of hardening Modified Orowan Model dcdc Main parameters  : Stacking Fault Energy,  d One adjustable parameter Number of dislocations in the pile-up dose  Defaulting No Defaulting

DEN - Saclay – Nuclear Material departement "Influence of atomic displacement rate… " Workshop - Ulyanovsk - October 5, 2005 Ph. Dubuisson - 18 DMN / SRMA Modified "Orowan" model Hardening model permitting defaulting of Frank Loops  Saturation of Hardening Well description of experimental data by the model 330°CGood description of experimental data 375°CNeed data at high dose to verify Voids  data / Model in SA 304  2 steels :  0  Voids  cm -2  26 Jm -2  cm -2  42 Jm -2  CW °C 375°C 330°C 375°C SA 304L 

DEN - Saclay – Nuclear Material departement "Influence of atomic displacement rate… " Workshop - Ulyanovsk - October 5, 2005 Ph. Dubuisson - 19 DMN / SRMA Slow Strain Rate Tests (SSRT) PWR environment T e = 320°C s °C 13 MPa O 2 ~ 0 ppb H 2 29 – 30 ml/kg <10 ppb 5 dpa SA 304L CW 316 T.E. of SSRT specimensstrongly 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 Flow rate 2 autoclave vol./h SA 304L CW 316  MPa Air PWR Air PWR Slight effect of He content

DEN - Saclay – Nuclear Material departement "Influence of atomic displacement rate… " Workshop - Ulyanovsk - October 5, 2005 Ph. Dubuisson - 20 DMN / SRMA Fracture surfaces ductile fracture IG fracture ductile fracture TG / IG fracture He (appm) % brittle fracture33,371,138,050,8 facet initiationTGIGTG»IGIG fracture in facetsTG»IGIG>TGTG=IGIG>TG 5 facets 3 facets Transgranular Intergranular SA 304L CW 316 high He Low He

DEN - Saclay – Nuclear Material departement "Influence of atomic displacement rate… " Workshop - Ulyanovsk - October 5, 2005 Ph. Dubuisson - 21 DMN / SRMA Conclusions - Perspectives  Tensile properties  Saturation dose at 5 dpa for SA 304L 10 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 loops No real saturation of loop number density and diameter  Hardening Model ●Orowan ModelNo saturation of hardening ●Modified Orowan Model permitting the defaulting of Frank loops Saturation of Hardening No evolution between 10 and 125 dpa

DEN - Saclay – Nuclear Material departement "Influence of atomic displacement rate… " Workshop - Ulyanovsk - October 5, 2005 Ph. Dubuisson - 22 DMN / SRMA 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 ?

DEN - Saclay – Nuclear Material departement "Influence of atomic displacement rate… " Workshop - Ulyanovsk - October 5, 2005 Ph. Dubuisson - 23 DMN / SRMA This work was performed through a collaboration between EDF, 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).