Experimental study of DHC. cladding and implications. dry storage conditions

Transcription

1 17th ASTM International Symposium Zirconium in the Nuclear Industry February, 2013, Hyderabad, India Experimental study of DHC of unirradiated d and irradiated d fuel cladding and implications to in-pile il operation and dry storage conditions V. Grigoriev 1, A-M. Alvarez-Holston 1, G. Lysell 1, D. Schrire 2, L. Hallstadius 3, S. T. Mahmood 4, and I. Arimescu 5 1 Studsvik Nuclear AB, Nyköping, Sweden 2 Vattenfall Nuclear Fuel AB, Stockholm, Sweden 3 Westinghouse Electric Sweden AB, Västerås, Sweden. 4 GNF-A, Vallecitos Nuclear Center, Sunol CA, USA 5 AREVA NP Inc., Richland WA, USA

2 Outline 1. Introduction 2. Tested materials and experimental details 3. Hydrides and DHC 4. Temperature dependence of V DHC 5. Effect of material strength on V DHC 6. Threshold stress intensity, K IH 7. Effect of hydrogen on V DHC and K IH 8. As a discussion: DHC in dry storage? 9. Summary 2

3 1.1 Introduction Zirconium alloys are susceptible to Delayed Hydride Cracking (DHC) It was recognized that DHC is responsible for stable crack propagation in pressure tubes (CANDU, RBMK) and, in some cases, in LWR fuel cladding Long axial splits in BWR cladding involve DHC initiated either from cracked hydride blisters or from PCI cracks DHC-type of outside-in in penetrating primary cracking have been observed in some in-reactor ramp tests What is the evidence that this is DHC? 3

4 1.2 Introduction Experience on DHC in pressure tubes has to be verified under conditions specific for fuel cladding (tubing geometry, alloying compositions, microstructure, texture, much higher hydrogen concentrations, presence of radial hydrides) Objectives of present work: - characteristics of DHC for irradiated fuel cladding (V DHC ; K IH ) - effect of hydrides and material strength on DHC - comparison with published data on V DHC (pressure tubes; cladding) 4

5 1.3 Introduction Most of experiments in present work have been performed within the frame of Studsvik Cladding Integrity Project, which is an international OECD/NEA Program. Close co-operation with IAEA Co-ordinated Research Project on Hydrogen and Hydride Degradation of Mechanical and Physical Properties of Zirconium Alloys. 5

6 6 2.1 Tested materials Irradiated cladding ( 35 tests; hydrogen ppm; different reactors ): - BWR cladding Zircaloy-2 ( LK1; LK2; LK2/L; LK2+/L; LK3; LK3/L tested for V DHC and K IH at 250 C and 300 C ) - PWR cladding ZIRLO ( tested for V DHC at temperatures from 130 C to 380 C and for K IH at 250 C) Unirradiated cladding ( >20 tests; hydrogen 200 ppm ): - standard SRA Zircaloy-4 (lot 86080) - CW Zircaloy-4 (lot 83786) - standard SRA ZIRLO SRA and CW Zircaloy-4 were also supplied to the IAEA CRP for PLT-DHC testing in 9-10 different laboratories 6

7 7 2.2 Material strength (ring tensile tests) Unirradiated cladding: plain rings having width of 2 mm were monotonically loaded in tension up to fracture at different temperatures. The strength (UTS) was calculated as a maximum load divided id d by specimen initial iti crosssectional area. Irradiated d cladding: modified d ring tension tests t (3-parts device) were performed, as a rule, at room temperature and at 300 C. In all cases, UTS values at T DHC ( C) are interpolated from the ring tensile test results 7

8 8 2.3 Hydride re-orientation C Loaded MPa ( entire volume ) Furnace cooling (~1 C/min) Air cooling (~10 C/min) ( local area ) before after 8

9 9 2.4 DHC tests (axial crack propagation, PLT) T peak 1-2ºC/min T DHC 3-5ºC/min Loading K I > K IH 9

10 DHC tests (outside-in crack propagation) No V DHC measurements Incipient crack at cladding surface WT = 0.57 mm 10

11 3.1 Radial hydrides and radial DHC 11 In unirradiated cladding radial hydrides do not crack at 250 C. However, radial hydrides crack easily at RT creating an incipient crack. Radial DHC in unirradiated cladding could only be initiated in the presence of incipient crack. Chevron pattern indicates the directions of DHC crack propagation. 11

12 Radial hydrides and axial DHC Unirradiated cladding No obvious effect of radial hydrides on DHC velocity Lower cooling rate appears to provide slightly higher DHC velocity 12

13 Radial hydrides and axial DHC Irradiated cladding No obvious effect of radial hydrides on DHC velocity Perhaps, Perhaps lower cooling rates provide higher DHC velocities 13

14 4.1 Temperature dependence of V DHC 14 Specific parameters of temperature dependence: Q, max V DHC, UTL. Each material has its individual set of those parameters. 14

15 Comparison with pressure tubes V DHC data for irradiated d cladding fit well to published data on pressure tubes 15

16 4.1a Temperature dependence of V DHC 16 Two groups of data for evaluation of the effect of material strength: - V DHC at 250 C - Arrhenius part of V DHC (T C) 16

17 Effect of strength on V DHC at 250 C [10] = IAEA TECDOC-1649 (2010) Good correlation with published data on unirradiated LWR / CANDU cladding 17

18 Normalized V DHC versus strength [11] = Markelov et.al. (2011) Data within Arrhenius part of V DHC (T C) normalised with: - hydrogen diffusivity, D, (Sawatzky, 1960), -solubility limit it for dissolution of hydrogen, C, (Kearns, 1967).. 18

19 7.1 Effect of hydrogen on V DHC 19 Hydrogen concentration in irradiated fuel cladding does not appear to affect the V DHC. 19

20 7.2 Effect of hydrogen on K IH 20 Data for irradiated cladding (250 and 300ºC) For unirradiated SRA Zry-4 (250 C/200 ppm): K IH m = 8-15 MPa m and K IH s = 8 MPa m. More data are needed to confirm an effect of high hydrogen concentrations on K IH values. 20

21 Why overheating is needed in lab? There appears to be no principal difference if oversaturation of hydrogen in solid solution is reached by means of corrosion (in reactor) or by means of overheating (in laboratory). 21

22 DHC in dry storage? The conditions for spent nuclear fuel storage are quite different from those in reactor (neutron flux, cladding stress, temperature conditions). Existing practice in the storage (peak temperature of 400 C followed by slow cooling) is very similar to the thermal history of laboratory ato DHC tests. For DHC to occur: K I = σ (C a) > K IH, where the constant C = 1.2π/Q is based on the shape factor Q [26] =Tiffani (1965). At what temperature the DHC could start: at UTL (375 C)? or at TSSp (315 C)? DHC at UTL requires in dry storage an overheating above 400 C 22

23 23 Summary DHC in irradiated fuel cladding follows the expected trends based on DHC behavior in pressure tubes and unirradiated claddings Each material has its specific values of UTL, T max, and maximum V DHC. The UTL for irradiated cladding is about 375 C that is increased by 40 C compared to unirradiated condition. The DHC crack in material with higher strength is growing faster and cracking occurs up to higher temperatures (higher UTL). ) Experimental data suggest that neither amount of hydrides nor their orientation can drastically affect the V DHC. A few data indicate a probable increase of K IH values at high hydrogen concentrations. 23

24 24 Summary (continued) Based on comparison of available data on axial and radial V DHC in unirradiated CW Zircaloy cladding one can assume that axial V DHC values can be used for characterization of radial DHC, at least, as a first order estimation. One can assume that irradiated cladding in dry storage should be immune to the DHC until cladding temperature is close to TSSp for cladding, i.e. at C. 24

25 25 Acknowledgements The work was performed within the Studsvik Cladding Integrity Project (SCIP), which is an international OECD/NEA programme. The contribution and support from all participating members are greatly acknowledged. Sincere gratitude for many stimulating discussions is expressed to the team of IAEA Co-ordinated Research Project led by Dr. Kit Coleman. The personnel at Studsvik are gratefully acknowledged for their creativity and professionalism in performing all sample preparations, mechanical testing and post irradiation examinations. 25

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