Influence of Thermo-mechanical Cycling on Pre-hydrogenated Zircaloy-4 Embrittlement by Radial hydrides Jean Desquines*, Manfred Puls**, Stéphane Charbaut* and Marc Philippe* *IRSN **MPP Consulting IRSN/FRM-296 ind 5
Context & Objectives ▌ Situations where thermo-mechanical cycling of fuel rods is expected: - Dry cask storage, - Spent fuel transportation from one spent fuel pool to another one ▌ Expected influence of cycling based on literature data o No clear trend (Kearns, Mishima, Chu, Sakamoto, Billone,...) o Possibly more radial hydrides and rather stronger embrittlement ▌ Main Goals of the study: clarify the expected influence of cycling on radial hydride precipitation: - Relying on a large data set on single cycle test results using « C »-shaped samples, - Then cycle samples and evaluate the radial hydride precipitation and cladding embrittlement. 19 TH INTERNATIONAL SYMPOSIUM ON ZIRCONIUM IN THE NUCLEAR INDUSTRY 2
1- Experimental protocol 19 TH INTERNATIONAL SYMPOSIUM ON ZIRCONIUM IN THE NUCLEAR INDUSTRY 3
“C” -Shaped Compression Tests in the Elastic Range F 8 Load f qq = 0.0 f qq = 0.0 Stress free area 7 𝑀 s: curvilinear distance to A (mm) 6 5 4 Equator q A A 3 s x 1 CCT sample 0 2 r 1 Stress free area 0 0 0.1 0.2 0.3 0.4 0.5 0.6 Load 𝐺 𝑂 X : location within the cladding thickness 𝜏 𝑢 𝐵 = 0.013 𝑀 𝑛𝑛 (0: Outer Diameter, 1: Inner diameter) 𝜏 𝑢 𝑦, 𝑡 = 𝜏 𝑢 𝐵 . 𝑔 𝜄𝜄 𝑦, 𝑡 19 TH INTERNATIONAL SYMPOSIUM ON ZIRCONIUM IN THE NUCLEAR INDUSTRY 4
Radial Hydride Heat Treatments (RHT) considered F [H] contents – 50 to 236 wppm 350°C Max temperature RHT 400°C Max temperature RHT 1h 1h 2h F RHT 2h 400 F RHT 350 Compression load Temperature ( ° C) Compression load Temperature ( ° C) 365°C 300 315°C 300 20°C/h 200 200 20°C/h 100 100 5 N 5 N 0 0 0 4 8 12 16 20 0 4 8 12 16 20 Time (h) Time (h) • The load is applied as late as possible to limit the sample creep but significantly before hydride precipitation • The creep is expected to increase linearly with cycle number 19 TH INTERNATIONAL SYMPOSIUM ON ZIRCONIUM IN THE NUCLEAR INDUSTRY 5
Using CCT to study radial hydride precipitation Post-test metallography RHT, 1 to 30 cycles (Constant Load) F Hydrided sample Equator Failure test: CCT F. d at RT 19 TH INTERNATIONAL SYMPOSIUM ON ZIRCONIUM IN THE NUCLEAR INDUSTRY 6
Experimental device Fan cooling the load cell 500 N load cell Simple model of the test F. d Furnace L1 2.R e L2 Furnace Electromechanical tension- compression machine 19 TH INTERNATIONAL SYMPOSIUM ON ZIRCONIUM IN THE NUCLEAR INDUSTRY 7
Interpretation of the RHT load displacement record 𝑋𝑓𝑚𝑚 𝑙𝑜𝑝𝑥𝑜 𝑏𝑒𝑘𝑣𝑡𝑢𝑓𝑒 𝐺𝐹 𝑑𝑏𝑚𝑑𝑣𝑚𝑏𝑢𝑗𝑝𝑜 𝑛𝑏𝑢𝑓𝑠𝑗𝑏𝑚 𝑞𝑠𝑝𝑞𝑓𝑠𝑢𝑧 F. d 𝑑𝑠𝑓𝑓𝑞 + 𝜀 𝑠𝑝𝑒𝑡 𝑢ℎ𝑓𝑠𝑛𝑏𝑚 + 𝜀 𝑡𝑏𝑛𝑞𝑚𝑓 𝑓𝑚 𝑢ℎ𝑓𝑠𝑛𝑏𝑚 𝜀 = 𝜀 𝑝𝑔𝑔𝑡𝑓𝑢 + 𝜀 𝑡𝑏𝑛𝑞𝑚𝑓 + 𝜀 𝑡𝑏𝑛𝑞𝑚𝑓 Furnace REOR-103 – first cycle Reasonably good agreement is obtained between calculations and measurement assuming 0 creep 19 TH INTERNATIONAL SYMPOSIUM ON ZIRCONIUM IN THE NUCLEAR INDUSTRY 8
Interpretation of the RHT load displacement record F. d 𝑑𝑠𝑓𝑓𝑞 𝜀 𝑡𝑏𝑛𝑞𝑚𝑓 = 𝜀 − 𝜀 𝑛𝑝𝑒𝑓𝑚 REOR-103 – 30 RHT cycles Furnace 𝜀 − 𝜀 𝑛𝑝𝑒𝑓𝑚 (𝑛𝑛) 0.8 0.6 0.4 0.2 0.0 0 100 200 300 400 500 -0.2 time (h) -0.4 Creep displacement (temperature stabilized) Creep deformation remains below ~1% diameter change after 30 cycles 19 TH INTERNATIONAL SYMPOSIUM ON ZIRCONIUM IN THE NUCLEAR INDUSTRY 9
RHT & CCT Test matrix REOR-XXX ( [H](wppm) ) – Test ID and associated hydrogen content RHT parameters Italic: the hydrogen content of tested conditions in past programs . 1 2 5 10 30 Cycles Number 53 ; 63 ; 69 ; 74 ; REOR-118 ( 95 ) Max RHT temperature(°C) REOR-119 (86) REOR-117 ( 109 ) REOR-115 ( 101 ) REOR-116 (94) 350 127 ; 141 ; REOR-114 ( 176 ) REOR-113 (210) REOR-111 (138) REOR-110 ( 139 ) 177 ; 217 ; REOR-112 (209) REOR-106 ( 203 ) 309 ; 322 ; 525 ; 540 REOR-131 ( 61 ) REOR-126 ( 62 ) REOR-121 ( 75 ) REOR-133 ( 141 ) REOR-120 ( 78 ) REOR-124 ( 148 ) 400 REOR-134 ( 187 ) REOR-125 ( 147 ) REOR-135 ( 177 ) 217 ; 218 ; REOR-123 ( 236 ) 398 ; 478 350 & 400°C Max temperature – 1 to 30 cycles – [H] content below 240 wppm Maximum stress at equatorial location: between 190 and 220 MPa 19 TH INTERNATIONAL SYMPOSIUM ON ZIRCONIUM IN THE NUCLEAR INDUSTRY 10
2- Analysis of Metallographic data 19 TH INTERNATIONAL SYMPOSIUM ON ZIRCONIUM IN THE NUCLEAR INDUSTRY 11
RHT at 350°C max temperature Cycles 1 5 10 30 REOR-17 [11] REOR-106 REOR-112 REOR-113 [H] (wppm) Metallographs at equatorial location 200 REOR-14 [11] REOR-110 REOR-111 130 • When cycling: higher density but comparable fraction of radial hydrides • Saturated influence of cycling over 5 cycles 19 TH INTERNATIONAL SYMPOSIUM ON ZIRCONIUM IN THE NUCLEAR INDUSTRY 12
RHT at 400°C max temperature Cycles 1 5 10 [H] REOR-134 REOR-123 REOR-135 Metallographs at equatorial location (wppm) 200 REOR-133 REOR-125 REOR-124 150 REOR-120 REOR-121 REOR-131 75 At H~200 wppm: higher density but comparable fraction of radial hydrides Below H~150 wppm: no influence of cycling 19 TH INTERNATIONAL SYMPOSIUM ON ZIRCONIUM IN THE NUCLEAR INDUSTRY 13
Influence of cycling and stress on incipient radial hydride precipitation 350°C RHT 400°C RHT 120 100 Tangential stress (MPa) Mixed radial & circumferential hydrides 80 60 Temp. RHT cycles 40 1 2 350 5 20 10 Circumferential hydrides 30 0 0 100 200 300 400 500 600 [H] (wppm) Slight influence of cycling at 350°C maximum temperature No clear influence at 400°C 19 TH INTERNATIONAL SYMPOSIUM ON ZIRCONIUM IN THE NUCLEAR INDUSTRY 14
Influence of cycling and stress on radial hydride precipitation 100% Radial hydrides 0% RHT cycles 200 1 RHT maximum temperature (°C) (Eq.10) 180 2 Tangential stress (MPa) 350 5 160 10 140 30 1 120 5 400 10 Mixed radial & circumferential 100 450 1 hydrides 80 60 40 20 Circumferential hydrides 0 0 100 200 300 400 500 600 [H] (wppm) Limited influence of cycling on radial hydride precipitation stress thresholds 19 TH INTERNATIONAL SYMPOSIUM ON ZIRCONIUM IN THE NUCLEAR INDUSTRY 15
F, d 3- Radial hydride Embrittlement 19 TH INTERNATIONAL SYMPOSIUM ON ZIRCONIUM IN THE NUCLEAR INDUSTRY 16
Load displacement records – an illustration 20 crack propagation 18 16 Crack nucleation Brittle Plasticity 14 12 10 CCT-112 (209wppm) Elasticity 8 Crack nucleation 6 4 CCT-115 (101wppm) 2 0 0 0.5 1 1.5 2 2.5 Displacement (mm) Crack nucleation is associated with the first significant load-drop and deviation to the elastic-plastic trend 19 TH INTERNATIONAL SYMPOSIUM ON ZIRCONIUM IN THE NUCLEAR INDUSTRY 17
Load displacement records 20 REOR-135 REOR-112 18 REOR-106 REOR-110 Rather high [H] contents ductile 16 britlle &low fracture toughness Plasticity REOR-123 REOR-111 REOR-113 14 F/L(N/mm) REOR-121 12 REOR-119 REOR-118 REOR-124 Elasticity RCT-126 ductile 10 britle Rather low [H] contents REOR-134 REOR-114 &high fracture toughness REOR 133 8 REOR-120 REOR-125 6 REOR-116 REOR-115 REOR-117 propagation nucleation Crack Crack 4 2 REOR-131 0 0 0.5 1 1.5 2 2.5 Displacement (mm) Surprisingly ductile nucleation was associated with brittle crack propagation and brittle crack nucleation with ductile crack propagation 19 TH INTERNATIONAL SYMPOSIUM ON ZIRCONIUM IN THE NUCLEAR INDUSTRY 18
Crack nucleation displacement – 1 cycle Maximum RHT temperature Crack nucleation displacement 2.0 350°C 1.8 400°C 450°C 1.6 1.4 (mm) 1.2 1.0 0.8 0.6 0 100 200 300 400 500 600 [H] (wppm) Maximum sensitivity to crack nucleation at about [H]= 100 wppm, No obvious influence of maximum RHT temperature 19 TH INTERNATIONAL SYMPOSIUM ON ZIRCONIUM IN THE NUCLEAR INDUSTRY 19
Crack nucleation displacement – several cycles RHT cycles Temp. 350°C RHT cycling trend 2 Crack nucleation displacement 2.0 350 5 10 Protective influence of RHT cycling 30 5 1.5 400 (mm) 10 1.0 Detrimental influence of RHT cycling 0.5 0 100 200 300 400 500 600 [H] (wppm) • Regions associated with higher density of radial hydrides have a lower sensitivity to crack nucleation • Protective influence of RHT cycling on crack nucleation 19 TH INTERNATIONAL SYMPOSIUM ON ZIRCONIUM IN THE NUCLEAR INDUSTRY 20
Conclusion ▌ The thermo-mechanical cycling was studied relying on CCT tests ▌ Analysis of metallographs - At 350°C for H contents between 75 and 230 wppm, and at 400°C for H contents close to 200 wppm: shorter and denser hydrides observed, - A 400°C below 200 wppm, no influence of cycling is observed - The denser and short hydrides were associated with slightly easier radial hydride precipitation ▌ A protective influence of cycling was observed on radial hydride embrittlement. ▌ Extrapolation of these results to irradiated claddings: If there is any location containing 100 wppm, then no influence of cycling is expected 19 TH INTERNATIONAL SYMPOSIUM ON ZIRCONIUM IN THE NUCLEAR INDUSTRY 21
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