FOR LIFETIME EXTENSION»

Transcription

1 PAGE «ROLE OF MATERIALS FOR LIFETIME EXTENSION» Pascal YVON, Bernard MARINI and Benoit TANGUY Department of Materials for Nuclear Applications, CEA SACLAY

2 OUTLINE Context Effect of neutrons on materials Role of R&D for lifetime extension through two exemples Pressure vessel Internal structures Conclusions

3 SOME MATERIALS ISSUES FOR A PWR Primary circuit Defects under liner of ferritc steel (including vessel) Stress Corrosion Cracking of alloy 600 steam Thermal Fatigue of thermal barriers of primary pumps Thermal ageing embrittlement of some casted components(coudes, corps de pompes ) Thermal ageing embrittlement of some HAZ of ferritic steels with high P content Auxiliary Circuits Corrosion in dead zones Thermal and vibrational Fatigue Internals Irradiation damage of bolts (IA-SCC) Dimensional changes of internals structures under swelling Containement barriers rapid degradation of some concretes etc

4 IRRADIATED COMPONENTS C 10/15 dpa 5 6 years Fuel Assemblies Zr alloys Core Internals Nickel alloys ~ 320 C few 0.1 dpa years ~ 300 C 0.1 dpa years 155 bars 293 C Water H 2, LiOH, B Vessel Bainitic steel 16MND5 A508 Cl bars 328 C neutrons temperature mechanical stresses environment time Control rods Austenitic steels ~ 320 C ~ 10 dpa few years Core Internals Austenitic steels C dpa years

5 ROLE OF R&D AND METHODOLOGY R&D outputs of two kinds Existing materials : evaluation of ageing in order to perform preventive maintenance on replaceable components (internals) and justify the life extension of irreplaceable components (RPV) Future materials : understanding of ageing mechanisms to propose optimized materials or improve the design To perform R&D, we rely on Characterization (microstructural, mechanical,..) Simulation Modelling

6 EFFECT OF NEUTRONS Depending on their energies, neutron can have Nuclear effects (inelastic): - thermal neutrons Fission Capture (and subsequent nuclear reactions) 2 to 3 neutrons Fission 2 atoms : Short life radioactive fission products Capture + énergy (~ 200 Mev) neutron + heavy atom Ballistic effects (energy conservation) fast neutrons dpa point defects

7 EFFECTS OF ELASTIC COLLISIONS INSIDE A CRISTAL - For a transferred energy E t < E d (threshold energy typically ev) - > vibration of the crystal lattice -> heating - For a transferred energy E t > E d, the atom can be ejected from its atomic site and move through the crystal to other atomic sites (mean free path ~ several atomic sites) - This creates a vacancy + a self-interstitial atom. This is a Frenkel pair. PKA : Primary Knock on Atom

8 EFFECTS OF ELASTIC COLLISIONS INSIDE A CRISTAL - For a transferred energy E t < E d (threshold energy typically ev) -> vibration of the crystal lattice -> heating - For a transferred energy E t > E d, the atom can be ejected from its atomic site and move through the crystal to other atomic sites (mean free path ~ several atomic sites) - This creates a vacancy + a self-interstitial atom. This is a Frenkel pair. Vacancy Interstitial Lattice distorsion

9 DISPLACEMENT CASCADE (E T >> E D ) For a transferred energy large compared to E d, the ejected atom transfers part of its energy to other atoms of the crystal lattice these other atoms can then displace other atoms. The primary knock on atom induces a displacement cascade Vacancy : yellow Interstitials : red Displaced atoms : blue

10 Engineering stress (MPa) MACROSCOPIC EFFECTS OF IRRADIATION ON MATERIALS After irradiation evolution of mechanical properties can be observed For instance the tensile testing properties of steel, but also embrittlement, dimensional changes, enhanced corrosion, precipitation, segregation, amorphisation SA irradiated and tested at 325 C 9 dpa 5,5 dpa 3,5 dpa 2 dpa 1 dpa 400 0,8dpa 200 Unirradiated 0 0% 10% 20% 30% 40% 50% Engineering strain (%)

11 PWR VESSEL : SECOND SAFETY BARRIER Operating life limited by the vessel embrittlement Arrêté de Article 9: «le constructeur montrera en particulier que l'appareil ne présente aucun risque de rupture brutale en exploitation.»

12 PWR PRESSURE VESSEL Temperatures : C Coolant pressure: 155 bar = 4400 mm e = 220 mm Gross weight 450 t Steel A 508 Cl. 3 = 16 MND 5 (bainitic steel) Internal cladding: 304 L = Z 2 CN (austenitic stainless steel)

13 IRRADIATION EFFECTS Neutron irradiation embrittles the vessel material Ductile to brittle transition depends on material = n/cm², Tirr = 288 C = n/cm², Tirr = 288 C Lee, 2000

14 ASSESSMENT OF VESSEL INTEGRITY

15 K (MPa. m) INTEGRITY ASSESSMENT F M = K Ic / K > C s C S is depending on: the transient category the initiation mode (fragile / ductile) 200 K Ic (0 y.) Atoms for future October 24th 2012, Paris F M (0 y.) T ( C)

16 K (MPa. m) INTEGRITY ASSESSMENT F M = K Ic / K > C s C S is depending on: the transient category the initiation mode (fragile / ductile) 200 K Ic (0 y.) K Ic (x years) 150 TT Atoms for future October 24th 2012, Paris F M (0 y.) F M (x y.) T ( C)

17 K (MPa. m) INTEGRITY ASSESSMENT F M = K Ic / K > C s C S is depending on: the transient category the initiation mode (fragile / ductile) 200 K Ic (0 y.) K Ic (x years) 150 TT Effect of irradiation on DBDT depends on material 100 This dependance has to be known in order to know the life expectancy of the vessel 50 0 Atoms for future October 24th 2012, Paris F M (0 y.) F M (x y.) T ( C)

18 INTEGRITY ASSESSMENT The operator must be able to predict irradiation induced embrittlement in order to guarantee the absence of risk of sudden break Extrapolation is difficult given the empirical nature of embritittlement models based on PSI and MTR irradiations Enrichment of data base can lead to significant changes Short and mid termr&d: - Understanding of embrittlement phenomena by experiments and numerical simulations - Qualitative evaluation of critical parameters (chemical composition, neutron flux ) - Optimization of the empirical models Long term R&D: base embrittlement models on multiphysical and multi scale models.

19 True stress (MPa) MULTISCALE MODELLING Microstrutural representative mesh Microstructure Modeling Euro material A, irradiated C -50 C +25 C Crystal plasticity 0 0% 2% 4% 6% 8% 10% 12% 14% True strain Macroscopic behaviour Molecular Dynamics Ab initio Reference Atoms for future - October values 24th 2012, Paris Dislocation Dynamics Mechanisms: Dislocation Mobility Defect Strengths Local Rules Single crystal behaviour & crystal constitutive law

20 K (MPa m) EXAMPLE OF PREDICTION OF PWR VESSEL STEEL EMBRITTLEMENT UNDER NEUTRON IRRADIATION % 95% Master Curve 5% 1% Gamma = 0,0192 T= -150 C T= -125 C T= -100 C T= -75 C T= -50 C T= -25 C T= 0 C Température ( C)

21 K (MPa m) EXAMPLE OF PREDICTION OF PWR VESSEL STEEL EMBRITTLEMENT UNDER NEUTRON IRRADIATION % 95% Master Curve 5% 1% Gamma = 0,0135 T= -150 C T= -125 C T= -100 C T= -75 C T= -50 C T= -25 C T= 0 C Température ( C)

22 K (MPa m) EXAMPLE OF PREDICTION OF PWR VESSEL STEEL EMBRITTLEMENT UNDER NEUTRON IRRADIATION % 95% Master Curve 5% 1% Gamma = 0,0108 T= -125 C T= -100 C T= -75 C T= -50 C T= -25 C T= 0 C T= 25 C Température ( C)

23 K (MPa m) EXAMPLE OF PREDICTION OF PWR VESSEL STEEL EMBRITTLEMENT UNDER NEUTRON IRRADIATION % 95% Master Curve 5% 1% Gamma = 0,0099 T= -125 C T= -100 C T= -75 C T= -50 C T= -25 C T= 0 C T= 25 C Température ( C)

24 K (MPa m) EXAMPLE OF PREDICTION OF PWR VESSEL STEEL EMBRITTLEMENT UNDER NEUTRON IRRADIATION 99% 95% Master Curve 5% 1% Gamma = 0,0099 T= -100 C T= -75 C T= -50 C T= -25 C T= 0 C T= 25 C T= 50 C Température ( C)

25 K (MPa m) EXAMPLE OF PREDICTION OF PWR VESSEL STEEL EMBRITTLEMENT UNDER NEUTRON IRRADIATION 99% 95% Master Curve 5% 1% Gamma = 0,0113 T= -75 C T= -50 C T= -25 C T= 0 C T= 25 C T= 50 C T= 75 C Température ( C)

26 K (MPa m) EXAMPLE OF PREDICTION OF PWR VESSEL STEEL EMBRITTLEMENT UNDER NEUTRON IRRADIATION 99% 95% Master Curve 5% 1% Gamma = 0,0119 T= -50 C T= -25 C T= 0 C T= 25 C T= 50 C T= 75 C T= 100 C Température ( C)

27 JANNUS PLATFORM Triple beam chamber

28 Fe - 1% Mn MODEL ALLOY Fe 5+ (10 MeV) T = 400 C Flux: ions. m -2.s -1 Fluence: 1, ions.m -2 Number density: 2,0 0, m -3 Mean size: 21 nm 39 % of <001> and 61 % of <111> 200 nm 0.5 dpa 224 loops analysed z = -101, g = 020, bright field Number density: m -3 Radius: R= nm Composition: c Mn = % at 8 nm Volume:42x42x86,3 nm 3 Experimental evidence that radiation-induced segregation (under saturated alloy) can lead to formation of nanometre-scale solute clusters in ferritic alloys E. Meslin et al.

29 MODELLING OF CU PRECIPITATION BY KINETIC MONTE CARLO Fe or Cu On each crystal site Thermodynamics for interactions between species Kinetics, according to probability of occurrence Simulated time: One century

30 INTERNALS STRUCTURE OF PWR Design role of the Upper Internals : Align the rod control cluster assemblies with the fuel assemblies Immobilize the fuel assemblies PWR (155 bars) Upper Internals Design role of the Lower Internals: Support the core weight Circulation of the primary coolant Positioning of the core and fuel assemblies Protection of the RPV against irradiation embrittlement RPV The internals lifetime has an important impact on the nuclear power plant lifetime because the cost and difficulty of their replacement Fuel Assemblies Lower Internals PAGE

31 INTERNALS N4 (PWR 1350MW)

32 INTERNALS STRUCTURE OF PWR Lower Internals Baffle plates bolts / reactor vessel Former Baffle plates SS 304 4m Core Barrel 3m Material Former Bolts CW316 SS Choice of 304 and CW316 Austenitic Stainless steels for Internals Structures Bolts : mechanical ties between formers and baffle plates Chemical Composition (wt%): stainless steel C Si Mn P S Ni Cr Mo Fe / Bal Bal. PAGE

33 INTERNALS: DESIGN AND AGEING MECANISMS

34 IDENTIFIED DEGRADATION MECHANISMS

35 SAFETY ISSUES OF BAFFLE FORMER BOLTS CRACKING

36 IRRADIATION EFFECTS Consequences on mechanical properties and sensitivity to IASCC Dislocation loops HARDENING IRRADIATION CREEP LOCALIZATION-CHANNELLING TOUGHNESS DECREASE Segregations at grain boundaries : INCREASED SENSITIVITY TO SCC (IASCC)? Cavities, helium bubbles : POTENTIAL SWELLING

37 METHODOLOGY Internal structures of the PWR in austenitic stainless steels Better understanding at the micro/nano-scale Laboratory material selection and characterization Changes in microstructure, microchemistry and mechanical properties and material degradation -Swelling -Hardening -Irradiation creep -Loss of ductility -Susceptibility to IASCC, -Etc MPa R&D studies based upon the simulation of certain changes at temperatures used in PWRs Primary water C

38 Simulation tool : JANNUS CEA (irradiation with particles) Mutiscales modelling (rate theory, molecular dynamics, dislocations dynamics, crystal plasticity, mesoscopic mechanical behavior) OVERVIEW OF THE STUDIES RELATED TO INTERNALS AT CEA Microstructure and radiation hardening Radiation Induced segregation Localization of the deformation Swelling Irradiation Creep Experimental Neutron irradiated materials Mechanical tests TEM TEM-EDX, TAP Mechanical tests, TEM Irradiation at high doses, Swelling measurement, TEM In-reactor creep tests, TEM Modelling SCC of irradiated material In-reactor IASCC SCC tests on recirculation water loop Swelling mandrel

39 TESTS ON IRRADIATED STAINLESS STEELS Neutron Irradiations (volume) Link? PWR : dpa/s

40 IRRADIATION CREEP

41 IASCC Cracking of PWR bolts Dose : up to 80 dpa Temp: up to 370 C Correlation (temp, dose) / fissuration 25 m 200 m

42 IASCC - METHODOLOGY SCC tests on irradiated materials Sensitivity studies: SSRT tests (dynamic) Crack initiation studies: Constant load tests (~static) K1 hot cell, LECI (Nishioka, JNST,45,2008) Determination of a curve below which there is no crack initiation (depends on grade, environnement, temperature, )

43 IASCC - METHODOLOGY in situ SCC tests in représentative conditions Comparison of in pile and out of pile tests Final validation Crack initiation studies: Constant load tests (~static) (JMTR, Japan)

44 MECHANICAL BEHAVIOR OF IRRADIATED SS Microstructural evolution modification of mechanical behaviour Irradiation defects Black dots Frank loops Nano-voids Gas bubbles [Edwards et al., 2003; Pokor et al., 2004a] Radiation-induced segregation, second phase precipitation, etc. Formation of clear bands : localization of plastic deformation Evolution of mechanical properties [Nogaret, 2007] 304L SA 316 CW [Pokor, 2003]

45 MECHANICAL BEHAVIOR OF IRRADIATED SS Cristal scale modelling

46 MECHANICAL BEHAVIOR OF IRRADIATED SS Test en grands transformations avec 1000 grains cubiques Test sur un maillage d agrégat de 50 grains Behavior law for dense monocristal Monocristal Tests des lois cristallines implantées à l état nonirradié et irradié Simulation of mechanical behavior of unirradiated and irradiated materials Agrégat polycristallin simplifié Tests polycristallins et identification et validation des paramètres du modèle Test sur un monocristal poreux Identifier la loi d endommagement du monocristal Agrégat polycristallin avec tétraèdres de Voronoï Behavior law for porous monocristal Monocristal poreux Etudier la croissance et la coalescence d un monocristal poraux Homogenisation Agrégat polycristallin irradié homogénéisé Evaluation of swelling effect on mechanical properties

47 EXAMPLE OF MODELLING Microstructural informations (input of the modelling) Tensile curves as a function of irradiation (output of the modelling)

48 CONCLUSIONS The evolution of materials properties under irradiation control the component life time(component life time is determined by engineering approaches of the reactor safety ) Mechanisms behind these evolutions are numerous and complex Experiments, simulations and modelling approaches are simultaneously needed to study these mechanisms at the different scales. The main objective of R&D programs is to improve engineering approaches on existing material through qualitative understanding and macroscopic data production Long term R&D (basic research) is dedicated to mechanism understanding and their quantitative simulation through multiscale approach. Material design for future reactors is based on mechanisms understanding and modelling

49 TO LEARN MORE ABOUT MATERIALS

50 Thank you for your attention Commissariat à l énergie atomique et aux énergies alternatives Centre de Saclay Gif-sur-Yvette Cedex T. +33 (0) F. +33 (0) Direction DEN Départment DMNI Etablissement public à caractère industriel et commercial RCS Paris B