Environmentally assisted cracking in light water reactors. Semiannual report, July 1998-December 1998.

Abstract

This report summarizes work performed by Argonne National Laboratory on fatigue and environmentally assisted cracking (EAC) in light water reactors from July 1996 to December 1996. Topics that have been investigated include (a) fatigue of carbon, low-alloy, and austenitic stainless steels (SSs) used in reactor piping and pressure vessels, (b) irradiationassisted stress corrosion cracking of Type 304 SS, IC) EAC of Alloy 600, and (d) characterization of residual stresses in welds of boiling water reactor (BWR) core shrouds by numerical models. Fatigue tests were conducted on ferritic and austenitic SSs in water that contained various concentrations of dissolved oxygen to determine whether a slow strain rate applied during various portions of a tensile-loading cycle are equally effective in decreasing fatigue life. Slow-strain-rate-tensile tests were conducted in simulated BWR water at 288°C on SS specimens irradiated to a low fluence in the Halden reactor and the results were compared with similar data from a control-blade sheath and neutron-absorber tubes irradiated in BWRS to the same fluence level. Crack-growth-rate tests were conducted on compact-tension specimens from a low-carbon content heat of Alloy 600 in high-purity oxygenated water at 289°C. Residual stresses and stress intensity factors were calculated for BWR core shroud welds.iii 4. 5. 6. 7.8. 18.19. 20.21. 22.23. 24.25. 26. 27.28. 29.30. 31. 32.33. Plain carbon steels (CSs) and low-alloy steels (LASS) are used extensively in steam supply systems of pressurized and boiling water nuclear reactors (PWRS and BWRs) as piping and pressure vessel materials. Fatigue tests are being conducted on C S s and LASs in water and in air to establish the effects of material and loading variables on fatigue life. The results indicate a significant decrease in fatigue life when five conditions are satisfied simultaneously, viz.. when applied strain range, service temperature, dissolved-oxygen (DO) in the water, and s u l h r content of the steel are above a minimum threshold level, and the loading strain rate is below a threshold value. Only a moderate decrease in fatigue life is observed in light water reactor (LWR) environments when any one threshold condition is not satisfied. This report presents (a) results of fatigue tests that examine the influence of the reactor environment on crack initiation and crack growth of C S s and LASs, and (b) data on austenitic stainless steel (SS) that establish the effects of various loading and environmental variables on fatigue lives of these steels.Crack lengths as a function of fatigue cycles were determined for A533-Gr B LAS and A106-Gr B CS in air and water environments. The results indicate that, in air, fatigue cracks that are 10 p m or longer, form quite early during fatigue (<lo% of life) even at low strain ranges (~0.4%). The decrease in fatigue life of CS and LASs in high-DO water is primarily caused by the effects of environment on the growth of short cracks. During the initial stages of fatigue ...