Effects of LWR coolant environments on fatigue lives of austenitic stainless steels

The ASME Boiler and Pressure Vessel Code fatigue design curves for structural materials do not explicitly address the effects of reactor coolant environments on fatigue life. Recent test data indicate a significant decrease in fatigue life of pressure vessel and piping materials in light water reactor (LWR) environments. Fatigue tests have been conducted on Types 304 and 316NG stainless steel in air and LWR environments to evaluate the effects of various material and loading variables, e.g., steel type, strain rate, dissolved oxygen (DO) in water, and strain range, on fatigue lives of these steels. The results confirm the significant decrease in fatigue life in water. The environmentally assisted decrease in fatigue life depends both on strain rate and DO content in water. A decrease in strain rate from 0.4 to 0.004%/s decreases fatigue life by a factor of {approx} 8. However, unlike carbon and low-alloy steels, environmental effects are more pronounced in low-DO than in high-DO water. At {approx} 0.004%/s strain rate, reduction in fatigue life in water containing <10 ppb D is greater by a factor of {approx} 2 than in water containing {ge} 200 ppb DO. Experimental results have been compared with estimates of fatigue life based on the statistical model. The formation and growth of fatigue cracks in austenitic stainless steels in air and LWR environments are discussed

Environmentally assisted cracking in light water reactors. Semiannual report, April 1994--September 1994, Volume 19

This report summarizes work performed by Argonne National Laboratory (ANL) on fatigue and environmentally assisted cracking (EAC) in light water reactors from April to September 1994. Topics that have been investigated include (a) fatigue of carbon and low-alloy steel used in piping and reactor pressure vessels, (b) EAC of austenitic stainless steels (SSs) and Alloy 600, and (c) irradiation-assisted stress corrosion cracking (IASCC) of Type 304 SS. Fatigue tests have been conducted on A106-Gr B and A533-Gr B steels in oxygenated water to determine whether a slow strain rate applied during different portions of a tensile-loading cycle are equally effective in decreasing fatigue life. Crack growth data were obtained on fracture-mechanics specimens of SSs and Alloy 600 to investigate EAC in simulated boiling water reactor (BWR) and pressurized water reactor environments at 289{degrees}C. The data were compared with predictions from crack growth correlations developed at ANL for SSs in water and from rates in air from Section XI of the ASME Code. Microchemical changes in high- and commercial-purity Type 304 SS specimens from control-blade absorber tubes and a control-blade sheath from operating BWRs were studied by Auger electron spectroscopy and scanning electron microscopy to determine whether trace impurity elements may contribute to IASCC of these materials

Environmentally assisted cracking in light water reactors. Semiannual report July 1996--December 1996

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) irradiation-assisted stress corrosion cracking of Type 304 SS, (c) 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

Crack initiation and crack growth behavior of carbon and low-alloy steels

Section III of the ASME Boiler and Pressure Vessel Code specifies fatigue design curves for structural materials. These curves were based on tests of smooth polished specimens at room temperature in air. The effects of reactor coolant environments are not explicitly addressed by the Code design curves, but recent test data illustrate potentially significant effects of LWR coolant environments on the fatigue resistance of carbon and low-alloy steels. Under certain loading and environmental conditions, fatigue lives of test specimens may be a factor of {approx}70 shorter than in air. Results of fatigue tests that examine the influence of reactor environment on crack imitation and crack growth of carbon and low-alloy steels are presented. Crack lengths as a function of fatigue cycles were determined in air by a surface replication technique, and in water by block loading that leaves marks on the fracture surface. Decreases in fatigue life of low-alloy steels in high-dissolved-oxygen (DO) water are primarily caused by the effects of environment during early stages of fatigue damage, i.e., growth of short cracks 100 {micro}m, crack growth rates in high-DO water are higher than in air by one order of magnitude. The effects of LWR environments on growth of short cracks are discussed

Geothermal resources, Wilcox Group, Texas Gulf Coast

Results are presented of a regional study to identify areas where the Wilcox Group contains significant thicknesses of sandstone with subsurface temperatures higher than 300/sup 0/F. Eight of these geothermal fairways were identified. Control for this study was based on wells chosen so as to provide stratigraphic dip sections spaced 15 to 20 miles apart along the entire Texas Gulf Coast. Electrical well logs from the eight fairways are shown. (MHR

ii Effects of Thermal Aging on Fracture Toughness and Charpy–Impact Strength of Stainless Steel Pipe Welds by

The degradation of fracture toughness, tensile, and Charpy–impact properties of Type 308 stainless steel (SS) pipe welds due to thermal aging has been characterized at room temperature and 290°C. Thermal aging of SS welds results in moderate decreases in Charpy–impact strength and fracture toughness. For the various welds in this study, upper–shelf energy decreased by 50–80 J/cm 2. The decrease in fracture toughness J–R curve or JIC is relatively small. Thermal aging had little or no effect on the tensile strength of the welds. Fracture properties of SS welds are controlled by the distribution and morphology of second–phase particles. Failure occurs by the formation and growth of microvoids near hard inclusions; such processes are relatively insensitive to thermal aging. The ferrite phase has little or no effect on the fracture properties of the welds. Differences in fracture resistance of the welds arise from differences in the density and size of inclusions. Mechanical–property data from the present study are consistent with results from other investigations. The existing data have been used to establish minimum expected fracture properties for SS welds