Impact of the irradiation damage recovery during transportation on the subsequent room temperature tensile behavior of irradiated zirconium alloys

Transcription

1 Impact of the irradiation damage recovery during transportation on the subsequent room temperature tensile behavior of irradiated zirconium alloys B. Bourdiliau 1, F. Onimus 2, C. Cappelaere 1, V. Pivetaud 1, P. Bouffioux 3, V. Chabretou 4, A. Miquet 5 1 CEA-Saclay,DEN, Service d'etudes des Matériaux Irradiés, Gif-sur-Yvette, France 2 CEA-Saclay,DEN, Service de Recherches Métallurgiques Appliquées, Gif-sur-Yvette, France 3 EDF/R&D Les Renardières, Ecuelles, Moret sur Loing, France 4 AREVA, AREVA NP SAS, 10 Rue Juliette Récamier, Lyon, France 5 EDF/SEPTEN Avenue Dutriévoz, Villeurbanne, France 1

2 Industrial background : end of life of the fuel assembly Spent Nuclear Fuel Assembly In-reactor use Dry transportation In-pool storage Fuel assembly Neutron irradiation of the Zr alloy cladding Post-irradiation creep of the Zr alloy cladding and radiation damage recovery Room temperature mechanical properties at retrieval? 2

3 Aim of the study : Gain a better knowledge and understanding of the effects of transportation on the subsequent room temperature mechanical behavior of irradiated zirconium alloys Successive steps of the project : a long story Irradiation in PWR, then defueling Internal pressure creep tests at various temperatures and duration with various applied stress -> simulate the transportation Machining of a ring specimen out of the tested tube and tensile testing at room temperature + hardness tests after creep -> simulate the retrieval + comparison with non-irradiated and as-irradiated Zr alloy Thin foils taken out of the specimens and TEM examinations Understanding of the mechanical behavior at room temperature after transportation 3

4 Materials studied and testing conditions Nuclear Materials Department Two studied materials : -Stress relieved annealed (cold-worked) Zy-4 Chemical composition (% wt) %wt O Sn Fe Cr SRA Zy Nb 0 Zr Bal. -Recrystalization annealed Zr-1%Nb alloy RXA Zr-1%Nb Bal. Recrystallized Zr-1%Nb alloy -> Creep test conditions before tensile tests z Creep temperature ( C) Creep stress (MPa) Creep duration (hours) ZrNb-1 5 PWR cycles q ZrNb-2 4 PWR cycles µm ZrNb-3 4 PWR cycles ZrNb-4 4 PWR cycles As-irradiated ZrNb-5 Non-irradiated > Ring tensile tests performed after creep tests at room temperature, at a strain rate of s -1 4

5 Mechanical test results 5

6 Stress (MPa) Room temperature mechanical behavior after postirradiation creep 800 ZrNb-4 (as-irradiated) 700 ZrNb-2 (400 C MPa h) ZrNb-1 (400 C MPa - 3,301 h) 600 ZrNb-3 (420 C MPa h) ZrNb-5 (non irradiated) Nuclear Materials Department % 10% 20% 30% 40% 50% Strain (%) - Irradiation induced hardening (increase of the yield stress and ultimate tensile strength - Loss of macroscopic ductility (decrease of the uniform elongation), but the failure remains ductile (strong necking) - Recovery of the radiation induced hardening after creep test - Recovery of the macroscopic ductility of the material (uniform elongation) - But partial recovery of both the hardening and the macroscopic ductility 6

7 TEM examinations and discussion on the deformation mechanisms and mechanical behavior Stress (MPa) TEM TEM ZrNb-4 (as-irradiated) ZrNb-2 (400 C MPa h) ZrNb-1 (400 C MPa - 3,301 h) ZrNb-3 (420 C MPa h) ZrNb-5 (non irradiated) % 10% 20% 30% 40% 50% Strain (%) 7

8 Nuclear Materials Department TEM on the as-irradiated specimen after testing -> creation of a high density of small loops 10 nm 200 nm Before irradiation 50 nm After irradiation c a2 a3 Pyramidal P1 Prismatic P b=<c+a> a1 b=<a> Basal B Before irradiation -> homogeneous glide of dislocations in the prismatic planes mainly (at 20 C) 16th Ring tensile test at room temperature after irradiation Channels in the basal plane Channels in the prismatic planes After irradiation : - Heterogeneous deformation inside the grains (dislocation channeling) - both basal and prismatic slip can be activated (depending on the grain International Symposium, Zirconium in the Nuclear Industry 9-13 May Chengdu,slip Sichuan Province, China B. Bourdiliau, F. Onimus et al. 8 orientation) ->ASTM easier basal than before irradiation

9 Interpretation of the mechanical behavior of the as-irradiated specimen Channel Thin foil Channel Stress (MPa) ZrNb-4 (as-irradiated) ZrNb-2 (400 C MPa h) ZrNb-1 (400 C MPa - 3,301 h) ZrNb-3 (420 C MPa h) ZrNb-5 (non irradiated) 0% 10% 20% 30% 40% 50% Strain (%) High density of small loops -> pinning of dislocations -> radiation hardening Clearing of loops by gliding dislocations -> microscopic strain softening -> early localization of the deformation at the specimen scale -> decrease of the Uniform Elongation observed during ring tensile tests -> Why is the basal slip more easily activated after irradiation than before irradiation? 9

10 Effect of irradiation on the activated slip systems Nuclear Materials Department P B schematic CRSS (MPa) (arbitrary values) Prismatic CRSS before irradiation Prismatic CRSS after irradiation Basal CRSS before irradiation Basal CRSS after irradiation (Akthar, 1973) Temperature ( C) - Higher radiation hardening of the prismatic slip than on the basal slip -> due to differences in interactions between irradiation induced loops and dislocations gliding either in the prismatic planes or in the basal plane (see Poster Onimus et al.) -> This could explain that both the prismatic and the basal slip are activated after irradiation 10

11 Microstructure evolution during transportation As-irradiated Heat treatment at 400 C during 250 h Creep test at 400 C during 240 h under 130 MPa -> decrease of the loop density ( + increase of the loop size) (Ribis et al., ASTM 2007) 11

12 TEM after ring tensile test at room temperature following transportation Prismatic glide mainly Testing after creep at 400 C Testing after creep at 420 C After ring tensile test following post-irradiation creep : -> very few remaining loops -> homogeneous prismatic glide mainly, no channel 12

13 Interpretation of the effects of transportation on the mechanical behavior at room temperature After post-irradiation creep : ZrNb-4 (as-irradiated) ZrNb-2 (400 C MPa h) -> decrease of the loop density -> recovery of the radiation hardening Stress (MPa) ZrNb-1 (400 C MPa - 3,301 h) ZrNb-3 (420 C MPa h) ZrNb-5 (non irradiated) Homogeneous glide of dislocations, no dislocation channeling -> recovery of the uniform elongation 0 0% 10% 20% 30% 40% 50% Strain (%) But residual hardening and lower ductility than the non irradiated material -> What is the origin of this residual hardening? 13

14 Sources of residual hardening : the remaining loops Hardness (kg.mm -2 ) Remaining <a> loops after post-irradiation creep at 400 C during 240 h under 130 MPa Hardness after heat treatments and post-irradiation creep After creep C C Recovery C C Recovery C C Recovery C Stress (MPa) ZrNb-4 (as-irradiated) ZrNb-2 (400 C MPa h) ZrNb-1 (400 C MPa - 3,301 h) ZrNb-3 (420 C MPa h) ZrNb-5 (non irradiated) 0% 10% 20% 30% 40% 50% Strain (%) The remaining <a> loops could explain the observed residual hardening But even after a heat treatment at 450 C during 960 h a residual hardening is observed Non irradiated (no annealing) Time (h) (Ribis 16 th International et al., ASTM Symposium, 2007) Zirconium in the Nuclear Industry ASTM 9-13 May Chengdu, Sichuan Province, China B. Bourdiliau, F. Onimus et al. 14

15 Sources of residual hardening : the needle-like Nb precipitates 100 nm Hardness (kg.mm -2 ) Very low loop density after heat treatment at 450 C during 960 h Non irradiated (no annealing) After creep C C Recovery C C Recovery C C Recovery C Time (h) Enhanced precipitation by irradiation-> needle like Nb precipitates in Zr-Nb alloy After ring tensile test following post-irradiation creep -> dislocations pinned on Nb needles -> another source of residual hardening But after post-irradiation creep at 400 C during 900 h the hardness remains higher than after post-irradiation heat treatment at 400 C during 500 h -> Another source of residual hardening? 15

16 TEM after post-irradiation creep performed at 400 C Nuclear Materials Department Non irradiated material, after creep test at 400 C, 130 MPa, during 72 h (plastic strain <0.5%) After post-irradiation creep at 400 C, 130 MPa, during 240 h and 960 h (plastic strain <0.5%) (Ribis et al., ASTM 2007) -> Higher dislocation density after post-irradiation creep than for the non irradiated material (at same plastic strain) -> Enhanced dislocation multiplication during creep due to remaining loops and needle like Nb precipitates 16

17 TEM after ring test at room temperature following creep test Testing after creep at 400 C Testing after creep at 420 C After ring tensile test at room temperature following post-irradiation creep test -> High dislocation density that can be a source of residual hardening 17

18 Interpretation of the effects of transportation on the room temperature mechanical behavior (continued) The residual hardening is due to -The remaining loops -The needle-like Nb precipitates -The dislocations coming from the creep test Stress (MPa) ZrNb-4 (as-irradiated) ZrNb-2 (400 C MPa h) ZrNb-1 (400 C MPa - 3,301 h) ZrNb-3 (420 C MPa h) ZrNb-5 (non irradiated) % 10% 20% 30% 40% 50% Strain (%) Due to the higher dislocation density -> reduced work-hardening capability -> lower Uniform Elongation than the non irradiated material 18

19 Conclusions on the mechanical behavior at room temperature Effect of neutron irradiation : - radiation hardening due to the high density of small loops induced by irradiation - decrease of the uniform elongation due to the dislocation channeling mechanism - the failure occurs after a strong necking -> ductile failure - the evolution of the activated slip systems can be explained the differences in the radiation hardening of the basal slip and prismatic slip. Effects of transportation : - radiation hardening recovery due to the loop annealing during creep - recovery of the uniform elongation due to homogeneous glide of dislocations - but remaining hardening and lower ductility than the non irradiated material, can be explained by remaining loops, needle Nb precipitates, dislocations coming from the previous creep test -> reduced strain hardening capability -> lower uniform elongation -> better knowledge and understanding of the effects of transportation on the mechanical properties of the fuel assembly at retrieval 19

20 Thank you! 20