MECHANICAL BEHAVIOR AT HIGH TEMPERATURE OF HIGHLY OXYGEN- OR HYDROGEN- ENRICHED α AND (PRIOR-) β PHASES OF ZIRCONIUM ALLOYS

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1 MECHANICAL BEHAVIOR AT HIGH TEMPERATURE OF HIGHLY OXYGEN- OR HYDROGEN- ENRICHED α AND (PRIOR-) β PHASES OF ZIRCONIUM ALLOYS I. Turque 1,2, R. Chosson 1,2,3, M. Le Saux 1*, J.C. Brachet 1, V. Vandenberghe 1,4, J. Crépin 2, and A.F. Gourgues-Lorenzon 2 with financial contributions from 1 DEN-Service de Recherches Métallurgiques Appliquées (SRMA), CEA, Université Paris-Saclay, F Gif-sur-Yvette, France 2 MINES ParisTech, PSL Research University, Centre des matériaux, CNRS UMR 7633, BP 87, Evry, France 3 Now at AREVA NP, Lyon Cedex 06, France 4 Now at DEN-Service d Etudes Mécaniques et Thermiques (SEMT), CEA, Université Paris-Saclay, F Gif-sur-Yvette, France *Corresponding author, matthieu.lesaux@cea.fr 18 TH INTERNATIONAL SYMPOSIUM ON ZIRCONIUM IN THE NUCLEAR INDUSTRY, MAY 15-19, 2016, HILTON HEAD, SC, USA 23 MAI 2016 I. TURQUE ET AL., 18TH INTERNATIONAL SYMPOSIUM ON ZIRCONIUM IN THE NUCLEAR INDUSTRY PAGE 1

2 INTRODUCTION β Zr (cc) α Zr (hcp) β ZrO 2 (tetra.) α ZrO 2 (mono.) Loss Of Coolant Accident (LOCA) β Zr ZrO 2 ZrO 2 Steam Oxygen content ZrO 2 ~25 wt.% ~66 at.% α Zr (O) 2-7 wt.% at.% β Zr wt.% 1-5 at.% Zr O β Zr α Zr (O) Distance from the outer surface ZrO 2 α Zr (+ H) Temperature of the cladding at a given axial position ( C) Zr + 2H 2 O 2H 2 + ZrO 2 ANL, ICL#2 Steam ZrO 2 α Zr (O) Prior-β Zr CEA Ex.: Single-side oxidation <900s 23 MAI 2016 I. TURQUE ET AL., 18TH INTERNATIONAL SYMPOSIUM ON ZIRCONIUM IN THE NUCLEAR INDUSTRY PAGE 2 Time JAEA, A 3-1 Example of Large Break LOCA transient Prior-β Zr ZrO 2 α Zr (O)

3 INTRODUCTION Steam Hydrogen content ZrO 2 α Zr (O) β Zr H (β Zr -stabilizer element) Distance from the outer surface Secondary hydriding: local hydrogen concentrations up to ~ wt.ppm (21-27 at.%) Mechanical behavior and integrity of the oxidized cladding during and after LOCA-like thermal-mechanical transients? Mechanical behavior at high temperature of the α Zr (O) phase containing more than 2 wt.% of oxygen? Mechanical behavior at high and low temperature of the (prior-) β Zr phase containing up to 3000 wt.ppm of hydrogen? 23 MAI 2016 I. TURQUE ET AL., 18TH INTERNATIONAL SYMPOSIUM ON ZIRCONIUM IN THE NUCLEAR INDUSTRY PAGE 3

4 OXYGEN-ENRICHED α Zr PHASE Materials and experimental procedures M5 is a registered trademark of AREVA NP in the USA or other countries Oxidation 1100 C 2 1 Annealing 1200 C 3h Stress 2-31 MPa Creep test C * Prepared from M5 cladding tubes Temperature Strain Time Axial tension Steam O Secondary vacuum O Steady-state strain rate O content α Zr ZrO 2 α Zr (O) β Zr α Zr (O) β Zr α Zr (O) MAI 2016 I. TURQUE ET AL., 18TH INTERNATIONAL SYMPOSIUM ON ZIRCONIUM IN THE NUCLEAR INDUSTRY PAGE 4

5 OXYGEN-ENRICHED α Zr PHASE Microstructure of model materials Average oxygen contents: 2, 3.2, 4.3 and 5.8 wt.% Zirconia completely reduced and oxygen concentration rather homogeneous within the samples (with the exception of the one with 5.8 wt.% of O in average) Oxygen-enriched model materials mainly composed of coarse lamellae or large grains ( 100 µm) of α Zr (O) phase (0 to 15% of residual untransformed β Zr phase) Content (wt.%) Distance from the outer surface (µm) 4.3 wt.% in average O Nb (EPMA) (EPMA) Texture (EBSD, neutron diffraction) comparable to that measured for the α Zr (O) phase formed during oxidation in steam at HT Modelα Zr (O) phase reasonably representative of the α Zr (O) phase observed in claddings oxidized at HT I. TURQUE ET AL., 18TH INTERNATIONAL SYMPOSIUM ON ZIRCONIUM IN THE NUCLEAR INDUSTRY PAGE 5

6 OXYGEN-ENRICHED α Zr PHASE Strain rate (s -1 ) DUCTILE 1000 C 20 MPa Ductile-to-brittle transition BRITTLE Failure without significant strain Oxygen content (wt.%) Tests Extrapolated models Viscoplastic flow Two creep regimes, depending on stress level Model α Zr (O) material (this study) M5 (Kaddour et al., 2004) Zircaloy-4 (Kaddour et al., 2004) Strain rate (s -1 ) O content n 1 n 5 2 wt.% O 3.2 wt.% O Stress (MPa) Dislocation regime 1100 C M 1000 C 900 C 800 C M C Nabarro-Herring, Coble or Harper-Dorn mechanisms? Creep resistance increases with increasing the oxygen content (e.g. strain rate of α Zr (2 wt.% O) times slower than that of α Zr phase without O addition) 23 MAI 2016 I. TURQUE ET AL., 18TH INTERNATIONAL SYMPOSIUM ON ZIRCONIUM IN THE NUCLEAR INDUSTRY PAGE 6

7 OXYGEN-ENRICHED α Zr PHASE Viscoplastic flow Modeling: strain rate ε & A T Q RT = exp σ n exp temperature Strain regime A (K.MPa -n.s -1 ) stress n ( BC ) Q (kj.mol -1 ) oxygen content Linear Power-law O B Strain rate (s -1 ) wt.% O 3.2 wt.% O Stress (MPa) Experiment Model M C 1000 C 900 C 800 C M4 M1000 C 23 MAI 2016 I. TURQUE ET AL., 18TH INTERNATIONAL SYMPOSIUM ON ZIRCONIUM IN THE NUCLEAR INDUSTRY PAGE 7

8 OXYGEN-ENRICHED α Zr PHASE Viscoplastic flow Modeling: C 20 MPa strain rate ε & A T Q RT stress = exp σ n exp temperature oxygen content ( BC ) O Strain rate (s -1 ) DUCTILE Ductile-to-brittle transition Failure without significant strain BRITTLE α Zr (O) Strain regime A (K.MPa -n.s -1 ) n Q (kj.mol -1 ) Linear Power-law B Oxygen content (wt.%) Tests Models Model α Zr (O) material (this study) α Zr (O) (this study) M5 (Kaddour et al., 2004) Zircaloy-4 (Kaddour et al., 2004) Zircaloy-2 (Donaldson and Evans, 1981) Zircaloy-2 (Burton et al., 1979) Zircaloy-4 (Chow et al., 1982) 23 MAI 2016 I. TURQUE ET AL., 18TH INTERNATIONAL SYMPOSIUM ON ZIRCONIUM IN THE NUCLEAR INDUSTRY PAGE 8

9 Fracture MECHANICAL BEHAVIOR OF THE HIGHLY OXYGEN-ENRICHED α Zr PHASE 0,3 Ductile-to-brittle transition Axial strain (-) 0,2 0,1 * Conduit Fractured à rupture Stoppé Stopped before failure avant rupture 800 C 900 C 1000 C 1100 C 0, Oxygen content (wt.%) *: Post-mortem profilometry measurement 23 MAI 2016 I. TURQUE ET AL., 18TH INTERNATIONAL SYMPOSIUM ON ZIRCONIUM IN THE NUCLEAR INDUSTRY PAGE 9

10 HYDROGEN-ENRICHED (PRIOR-) β Zr PHASE Mechanical behavior at low and high temperature of the (prior-) β Zr phase containing up to 3000 wt.ppm of hydrogen? 23 MAI 2016 I. TURQUE ET AL., 18TH INTERNATIONAL SYMPOSIUM ON ZIRCONIUM IN THE NUCLEAR INDUSTRY PAGE 10

11 HYDROGEN-ENRICHED (PRIOR-) β Zr PHASE Materials and experimental procedures Prepared from low-tin Zircaloy-4 cladding tubes ~3200 wt.ppm of H Temperature Hydrogen charging ~ wt.ppm 800 C Time Ar + H Hydrogen content measured for each sample by using an inert gas fusion thermal conductivity technique (+ DSC, µ-erda and neutron radiography on a selected number of samples) Satisfactory homogeneity of hydrogen content within the samples 23 MAI 2016 I. TURQUE ET AL., 18TH INTERNATIONAL SYMPOSIUM ON ZIRCONIUM IN THE NUCLEAR INDUSTRY PAGE 11

12 HYDROGEN-ENRICHED (PRIOR-) β Zr PHASE Materials and experimental procedures Partitioning of chemical elements between β Zr and α Zr (prior-β Zr ) phases during cooling and β Zr to α Zr phase transformation O O α Zr β Zr Temperature Hydrogen charging ~ wt.ppm 800 C 3000 wt.ppm of H Ar + H Time Thermo-Calc + Zircobase calculation 23 MAI 2016 I. TURQUE ET AL., 18TH INTERNATIONAL SYMPOSIUM ON ZIRCONIUM IN THE NUCLEAR INDUSTRY PAGE 12

13 HYDROGEN-ENRICHED (PRIOR-) β Zr PHASE Materials and experimental procedures Proeutectoidα Zr (prior-β Zr ) depleted in H and enriched in O Prior-β Zr transformed below the eutectoid temperature enriched in H and depleted in O Temperature Hydrogen charging ~ wt.ppm 800 C ~3200 wt.ppm of H Time Ar + H 23 MAI 2016 I. TURQUE ET AL., 18TH INTERNATIONAL SYMPOSIUM ON ZIRCONIUM IN THE NUCLEAR INDUSTRY PAGE 13

14 HYDROGEN-ENRICHED (PRIOR-) β Zr PHASE Materials and experimental procedures Axial tension Temperature Heat-treatment in the β Zr phase domain up to ~1200 C Hydrogen charging ~ wt.ppm 800 C * Tensile test C Stress Strain 0.1 s -1 Fast strain rate in order to minimize metallurgical evolutions and oxidation of the material during the test (faster than the rates at which the cladding can be subjected during a LOCA transient) Time Ar + H Air Growth of a thin oxide layer (5-10 µm) during heating in order to prevent hydrogen desorption 23 MAI 2016 I. TURQUE ET AL., 18TH INTERNATIONAL SYMPOSIUM ON ZIRCONIUM IN THE NUCLEAR INDUSTRY PAGE 14

15 Fracture MECHANICAL BEHAVIOR OF THE HIGHLY HYDROGEN-ENRICHED (PRIOR-) β Zr PHASE Material embrittled by H contents of wt.ppm for temperatures below 500 C ductile fracture H content Macroscopically brittle brittle fracture zones surrounded by zones of ductile fracture Proeutectoidα Zr (prior-β Zr ) depleted in H and enriched in O Prior-β Zr transformed below the eutectoid temperature enriched in H and depleted in O 23 MAI 2016 I. TURQUE ET AL., 18TH INTERNATIONAL SYMPOSIUM ON ZIRCONIUM IN THE NUCLEAR INDUSTRY PAGE 15

16 HYDROGEN-ENRICHED (PRIOR-) β Zr PHASE Fracture Effect of H on macroscopic ductility diminishes when temperature increases and becomes negligible beyond 500 C Cross-section reduction at failure ~98% ductile fracture T H content Macroscopically brittle 23 MAI 2016 I. TURQUE ET AL., 18TH INTERNATIONAL SYMPOSIUM ON ZIRCONIUM IN THE NUCLEAR INDUSTRY PAGE 16

17 HYDROGEN-ENRICHED (PRIOR-) β Zr PHASE Plasticity When the behavior is macroscopically ductile, the flow stress of the (prior-) β Zr phase containing between 1700 and 3200 wt.ppm of H is, compared to the one of the non-hydrided material higher at 500 C and below Proeutectoidα Zr enriched in O and depleted in H Prior-β Zr transformed below the eutectoid temperature, depleted in O but containing a very large amount of H at least partially precipitated under the form of strengthening nano-hydrides Figure: Highly hydrided material Proeutectoidα Zr Brittle H content Brittle H content lower at 700 C β Zr with all H in solid solution 23 MAI 2016 I. TURQUE ET AL., 18TH INTERNATIONAL SYMPOSIUM ON ZIRCONIUM IN THE NUCLEAR INDUSTRY PAGE 17

18 CONCLUSIONS Oxygen and hydrogen are known to be the main parameters responsible for embrittlement of zirconium alloys Mechanical behavior between 800 and 1100 C of the α Zr (O) phase containing between 2 and 5.8 wt.% of oxygen? Creep resistance of the oxygen-enriched α Zr (O) material increases with increasing O content, significantly higher than the creep resistance of the as-received material without additional O Two creep regimes observed: power-law regime for stresses higher than 15 MPa and nearly linear regime for lower stresses (further investigation needed to identify the mechanisms that drive the linear creep regime) Model α Zr (O) materials ductile between 800 and 1100 C for O contents between 2 and 3.2 wt.%, brittle, even at 1100 C, for O contents higher than 4 wt.% 23 MAI 2016 I. TURQUE ET AL., 18TH INTERNATIONAL SYMPOSIUM ON ZIRCONIUM IN THE NUCLEAR INDUSTRY PAGE 18

19 CONCLUSIONS Mechanical behavior between 20 and 700 C of the (prior-) β Zr phase containing between 1700 and 3200 wt.ppm of hydrogen? Young s modulus and plastic isotropy not substantially modified by H Material embrittled by H contents of wt.ppm for temperatures below 500 C: macroscopically brittle at 135 C and below for average H content of ~2000 wt.ppm and at C for ~3000 wt.ppm of H (further work to be done to determine the underlying mechanisms responsible for the effects of high H contents) Effect of H on macroscopic ductility diminishes when temperature increases and becomes negligible beyond 500 C When the behavior is macroscopically ductile, the flow stress of the material containing between 1700 and 3200 wt.ppm of H is higher than the one of the nonhydrided material at 500 C and below, and lower at 700 C 23 MAI 2016 I. TURQUE ET AL., 18TH INTERNATIONAL SYMPOSIUM ON ZIRCONIUM IN THE NUCLEAR INDUSTRY PAGE 19

20 Thank you for your attention Acknowledgments: D. Hamon, V. Lezaud, E. Rouesne, S. Urvoy, C. Toffolon-Masclet, P. Bonnaillie, M.H. Mathon, C. Raepsaet, G. Bayon J. Heurtel, A. Laurent, J.D. Bartout, A. Meddour, A. Koster, J.C. Teissedre PAGE 20 I. TURQUE ET AL., 18TH INTERNATIONAL SYMPOSIUM ON ZIRCONIUM IN THE NUCLEAR INDUSTRY Commissariat à l énergie atomique et aux énergies alternatives Centre de Saclay Gif-sur-Yvette Cedex T. +33 (0) F. +33 (0) Direction de l Energie Nucléaire Département des Matériaux pour le Nucléaire Service de Recherches Métallurgiques Appliquées 23 MAI 2016 Etablissement public à caractère industriel et commercial RCS Paris B

21 Back-up slides 23 MAI 2016 I. TURQUE ET AL., 18TH INTERNATIONAL SYMPOSIUM ON ZIRCONIUM IN THE NUCLEAR INDUSTRY PAGE 21

22 OXYGEN-ENRICHED α Zr PHASE Microstructure of model materials Average oxygen contents: 2, 3.2, 4.3 and 5.8 wt.% Zirconia completely reduced and oxygen concentration rather homogeneous within the samples (with the exception of the one with 5.8 wt.% of O in average) Oxygen-enriched model materials mainly composed of α Zr (O) grains, enriched in O and depleted in Nb and Fe (0 to 15% of residual untransformed β Zr phase) Coarse lamellae or large grains ( 100 µm) ZrO 2 α Zr (O) β Zr Content (wt.%) Content (wt.%) β Zr phase Distance from the outer surface (µm) 2 wt.% in average O Nb (EPMA) (EPMA) 4.3 wt.% in average O (EPMA) 0 Zircaloy-4, oxidized 600s in steam at 1100 C MAI 2016 I. TURQUE ET AL., 18TH INTERNATIONAL SYMPOSIUM ON ZIRCONIUM IN THE NUCLEAR INDUSTRY PAGE 22 Distance from the outer surface (µm) Nb (EPMA)

23 OXYGEN-ENRICHED α Zr PHASE Microstructure of model materials Residual β Zr phase: volume fraction between 0 and 15% 1,0 α Zr (O) phase volume fraction 0,8 0,6 0,4 0,2 0, C 1100 C 1000 C 1200 C 1100 C 1000 C Model α Zr (O) material Measurements Zr-1wt.%Nb-Fe-O Calculations Oxygen content (wt.%) 23 MAI 2016 I. TURQUE ET AL., 18TH INTERNATIONAL SYMPOSIUM ON ZIRCONIUM IN THE NUCLEAR INDUSTRY PAGE 23

24 OXYGEN-ENRICHED α Zr PHASE Microstructure of model materials Texture (EBSD, neutron diffraction) comparable to that measured for the α Zr (O) phase formed during oxidation in steam at HT Model α Zr (O) phase with 2 wt.% of O in average α Zr (O) layer formed during oxidation in steam at 1100 C on a M5 cladding tube 23 MAI 2016 I. TURQUE ET AL., 18TH INTERNATIONAL SYMPOSIUM ON ZIRCONIUM IN THE NUCLEAR INDUSTRY PAGE 24

25 OXYGEN-ENRICHED α Zr PHASE Viscoplastic flow C 2 MPa C 20 MPa Strain rate (s -1 ) DUCTILE Ductile-to-brittle transition BRITTLE Failure without significant strain Oxygen content (wt.%) Strain rate (s -1 ) DUCTILE Ductile-to-brittle transition BRITTLE Failure without significant strain Oxygen content (wt.%) Tests Models Model α Zr (O) material (this study) α Zr (O) (this study) M5 (Kaddour et al., 2004) Zircaloy-4 (Kaddour et al., 2004) Zircaloy-2 (Donaldson and Evans, 1981) Zircaloy-2 (Burton et al., 1979) Zircaloy-4 (Chow et al., 1982) For low stress levels, predictions of the present model deviate significantly from those of models from the literature, due to the introduction of a linear creep regime, observed for the first time in a highly O-enriched material 23 MAI 2016 I. TURQUE ET AL., 18TH INTERNATIONAL SYMPOSIUM ON ZIRCONIUM IN THE NUCLEAR INDUSTRY PAGE 25

26 OXYGEN-ENRICHED α Zr PHASE Viscoplastic flow Effect of the residual β Zr phase (up to 15%) estimated for the model material containing 2 wt.% of O in average, tested above 1000 C, by using a homogenization approach with a Taylor assumption Effect relatively small & & = Ε = ε α Zr O ) ε β Zr & Σ = 1 f σ O) + f σ β ( ) ( βzr αzr ( βzr Zr Temperature ( C) β Zr phase fraction < Oxygen content in α Zr (O) Softening due to β Zr phase Strengthening due to enrichment in oxygen in the α Zr (O) phase Linear < Power-law < Power-law < Multiplying factors on the creep rate 23 MAI 2016 I. TURQUE ET AL., 18TH INTERNATIONAL SYMPOSIUM ON ZIRCONIUM IN THE NUCLEAR INDUSTRY PAGE 26

27 HYDROGEN-ENRICHED (PRIOR-) β Zr PHASE Elasticity and plastic isotropy Young s modulus and plastic isotropy not significantly modified by the presence of high H contents 2D digital image correlation Young s modulus and plastic strain anisotropy 23 MAI 2016 I. TURQUE ET AL., 18TH INTERNATIONAL SYMPOSIUM ON ZIRCONIUM IN THE NUCLEAR INDUSTRY PAGE 27

28 HYDROGEN-ENRICHED (PRIOR-) β Zr PHASE Stress-strain curves 23 MAI 2016 I. TURQUE ET AL., 18TH INTERNATIONAL SYMPOSIUM ON ZIRCONIUM IN THE NUCLEAR INDUSTRY PAGE 28

29 HYDROGEN-ENRICHED (PRIOR-) β Zr PHASE Yield stress and reduction of area at fracture 23 MAI 2016 I. TURQUE ET AL., 18TH INTERNATIONAL SYMPOSIUM ON ZIRCONIUM IN THE NUCLEAR INDUSTRY PAGE 29

30 HYDROGEN-ENRICHED (PRIOR-) β Zr PHASE Uniform elongation Plastic strain localization occurs sooner in the highly hydrided material, when significant plastic strain occurs before failure 23 MAI 2016 I. TURQUE ET AL., 18TH INTERNATIONAL SYMPOSIUM ON ZIRCONIUM IN THE NUCLEAR INDUSTRY PAGE 30

31 HYDROGEN-ENRICHED (PRIOR-) β Zr PHASE Ductile-to-Brittle transition 23 MAI 2016 I. TURQUE ET AL., 18TH INTERNATIONAL SYMPOSIUM ON ZIRCONIUM IN THE NUCLEAR INDUSTRY PAGE 31