Influence of oxide films on primary water stress corrosion cracking initiation of alloy 600
Source abstract
Abstract
In the present study alloy 600 was tested in simulated pressurised water reactor (PWR) primary water, at 360 °C, under an hydrogen partial pressure of 30 kPa. These testing conditions correspond to the maximum sensitivity of alloy 600 to crack initiation. The resulting oxidised structures (corrosion scale and underlying metal) were characterised. A chromium rich oxide layer was revealed, the underlying metal being chromium depleted. In addition, analysis of the chemical composition of the metal close to the oxide scale had allowed to detect oxygen under the oxide scale and particularly in a triple grain boundary. Implication of such a finding on the crack initiation of alloy 600 is discussed. Significant diminution of the crack initiation time was observed for sample oxidised before stress corrosion tests. In view of these results, a mechanism for stress corrosion crack initiation of alloy 600 in PWR primary water was proposed.
Source: openaire · Quality: candidate unverified · Source record ↗
Research context
OpenAlex topics: Nuclear Materials and Properties · Hydrogen embrittlement and corrosion behaviors in metals · High-Temperature Coating Behaviors
SciVX discovery tags are terminology-based. OpenAlex topics are source-provided classifications.
References (30)
Show source-provided references
1. T. Magnin, F. Foct, O. De Bouvier, Hydrogen effects on PWRSCC mechanisms in monocrystalline and polycrystalline alloy 600, In: Proc. of Ninth Int. Symp. On Environmental Degradation of Materials in Nuclear Power Systems-Water Reactors, Newport Beach (1999) 27.10.1002/9781118787618.ch3 ↗
2. Oxidation-deformation interactions and effect of environment on the crack growth resistance of Ni-base superalloys
3. Mechanisms of environmentally enhanced deformation and intergranular cracking of Ni–16Cr–9Fe alloys
4. T. Cassagne, B. Fleury, F. Vaillant, O. De Bouvier, P. Combrade, An update on the influence of hydrogen on the PWSCC of alloy 600 in high temperature water, In: Proc. of Eighth Int. Symp. On Environmental Degradation of Materials in Nuclear Systems-Water Reactors, La Grange Park, Illinois (1997) 307.
5. R. Gate, B. Rassineux, A. Gelpi, M. Le Calvar, B. Sala, Mechanical and Thermohydraulic Testing on Steam Generator Tubing Cracked in Laboratory Stress Corrosion, In: Proc. of Progress in the Understanding and Prevention of Corrosion, Barcelona, Spain (1993) 1542.
6. Catlin · Automated Methods for Fracture and Fatigue Crack Growth ASTM STP · 1985 · 877 · 6710.1520/STP32880S ↗
7. J. Panter, B. Viguier, E. Andrieu, Microstructure evolution of alloy 600 during dry oxidation, In: Proc. of Les Embiez 2000, Les Embiez, Trans Tech Publications (2001) 141.10.4028/www.scientific.net/MSF.369-372.141 ↗
8. Monceau · J. Mater. Res. · 2000 · 15 · 66510.1557/JMR.2000.0099 ↗
9. C. Soustelle, M. Foucault, P. Combrade, K. Wolski, T. Magnin, PWSCC of Alloy 600: A Parametric Study of Surface Film Effects, In: Proc. of Ninth Int. Symp. on Environmental Degradation of Material in Nuclear Power Systems-Water Reactors (1999) 105.10.1002/9781118787618.ch10 ↗
10. Carette · Surf. Interface Anal. · 2002 · 24 · 13510.1002/sia.1269 ↗
11. S. Gardey, Etude de la corrosion généralisée des alliages 600, 690 et 800 en milieu primaire. Contribution à la compréhension des mécanismes, PhD Thesis, Université Pierre et Marie Curie (1998).
12. Giggins · Trans. Met. Soc. AIME · 1969 · 245 · 2495
13. Robertson · Corros. Sci. · 1991 · 32 · 44310.1016/0010-938X(91)90125-9 ↗
14. E.P. Simonen, L.E. Thomas, S.M. Bruemmer, Diffusion kinetic issues during intergranular corrosion of Ni-base alloys, In: Proc. of Corrosion 2000 (2000) 226.
15. T.S. Gendron, S.J. Bushby, R.D. Cleland, R.C. Newman, Oxidation embrittlement of alloy 600 in hydrogenated steam at 400 °C, In: Proc. of Eurocorr’96 (1997) M24.
16. Bricknell · Scr. Metall. · 1982 · 30 · 257
17. Iacocca · Metall. Trans. A · 1988 · 19 · 230510.1007/BF02645054 ↗
18. P.M. Scott, An overview of internal oxidation as a possible explanation of intergranular SCC of alloy 600 in PWRS, In: Proc. of Ninth Int. Symp. On Environmental Degradation of Materials in Nuclear Power Systems-Water Reactors, Newport Beach (1999) 3.10.1002/9781118787618.ch1 ↗
19. R.W. Staehle, Z. Fang, Comments on a proposed mechanism of internal oxidation for alloy 600 as applied to low potential SCC, In: Proc. of Ninth Int. Symp. on Environmental Degradation of Material in Nuclear Power Systems-Water Reactors (1999) 69.10.1002/9781118787618.ch7 ↗
20. On the role of oxygen in SCC as a function of temperature
21. T. Magnin, A unified model for trans and intergranular stress corrosion cracking in fcc ductile alloys, in: T. Magnin, Gras, J.M. (Eds.), Corrosion-deformation interactions, Fontainebleau, France, Les éditions de physique (1992) 410.
22. P. Hancock, Influence of vacancies produced by oxidation on the mechanical properties of nickel and nickel–chromium alloys, In: R.E. Smallman (Ed.), Vacancies ’76 Met. Soc. (1976) 215.
23. Bricknell · Metall. Mater. Trans. A · 1981 · 12 · 42510.1007/BF02648539 ↗
24. C.D. Thompson, H.T. Krasodomski, N. Lewis, G.L. Makar, Prediction of PWSCC in nickel base alloys using crack growth rate models, In: Proc. of Seventh Int. Symp. On Environmental Degradation of Materials in Nuclear Power Systems-Water Reactors (1995) 867.
25. Perusin · Acta Mater. · 2004 · 52 · 537510.1016/j.actamat.2004.07.043 ↗
26. Perusin · Mater. Sci. Forum · 2004 · 461–464 · 12310.4028/www.scientific.net/MSF.461-464.123 ↗
27. Johnson · Phys. Rev. B · 1976 · 13 · 436410.1103/PhysRevB.13.4364 ↗
28. Berge · Corros. Sci. · 1970 · 10 · 18510.1016/S0010-938X(70)80077-4 ↗
29. T.S. Gendron, P.M. Scott, S.M. Bruemmer, L.E. Thomas, Internal oxidation as a mechanism for steam generator tube degradation, In: Proc. of Third Int. Steam Generator and Heat Exchanger Conference, Toronto, Canadian Nuclear Society (1998) 5.18.
30. Bandy · Corrosion-Nace · 1984 · 40 · 42510.5006/1.3593950 ↗