Crack growth rates and fracture toughness of irradiated austenitic stainless steels in BWR environments.

Abstract

In light water reactors, austenitic stainless steels (SSs) are used extensively as structural alloys in reactor core internal components because of their high strength, ductility, and fracture toughness. However, exposure to high levels of neutron irradiation for extended periods degrades the fracture properties of these steels by changing the material microstructure (e.g., radiation hardening) and microchemistry (e.g., radiation-induced segregation). Experimental data are presented on the fracture toughness and crack growth rates (CGRs) of wrought and cast austenitic SSs, including weld heataffected-zone materials, that were irradiated to fluence levels as high as ≈ 2 x 10 21 n/cm 2 (E > 1 MeV) (≈ 3 dpa) in a boiling heavy water reactor at 288-300°C. The results are compared with the data available in the literature. The effects of material composition, irradiation dose, and water chemistry on CGRs under cyclic and stress corrosion cracking conditions were determined. A superposition model was used to represent the cyclic CGRs of austenitic SSs. The effects of neutron irradiation on the fracture toughness of these steels, as well as the effects of material and irradiation conditions and test temperature, have been evaluated. A fracture toughness trend curve that bounds the existing data has been defined. The synergistic effects of thermal and radiation embrittlement of cast austenitic SS internal components have also been evaluated. Paperwork Reduction Act StatementThis NUREG does not contain information collection requirements and, therefore, is not subject to the requirements of the Paperwork Reduction Act of 1995 (44 U.S.C. 3501 et seq.). Public Protection NotificationThe NRC may not conduct or sponsor, and a person is not required to respond to, a request for information or an information collection requirement unless the requesting document displays a current valid OMB control number. ForewordThis report presents the results of a study of simulated light-water reactor coolants, material chemistry, and irradiation damage and their effects on the susceptibility to stress-corrosion cracking of various commercially available and laboratory-melted stainless steels. This report is one of a series dating back about 8 years, describing such results, which are required to support analysis of the structural integrity of reactor internal components, many of which are subject to irradiation-assisted stress-corrosion cracking (IASCC).The earlier reports detailed crack growth rates in heat-affected zones adjacent to stainless steel weldments, and they comprised the final publications based on specimens irradiated in Phase I (of two) in the Halden test reactor. Phase I irradiations principally involved stainless steels of wide-ranging chemistry (including commercial steels of typical chemistry) and conventional heat treatment and product form processing. By contrast, this report is the first to present data from specimens irradiated in Phase II, which featured a variety of innovatively fabricated and engineered ...

Thermal Aging Phenomena in Cast Duplex Stainless Steels

“…As discussed above, the aging embrittlement depends on many materials and environmental parameters, including chemistry, processing condition, aging temperature, and irradiation condition. [1][2][3][4][5][6][7][37][38][39][40] Practically, the wide range of impact energy data influenced by the parameters mentioned above may not allow an integration into single figure in a discernible way. Instead, a schematic illustration was constructed to express a summary of the reviewed context above: Fig.…”

“…It is also well known that thermal aging has a synergistic effect with irradiation when the CASS components are used in or near a reactor. [37][38][39][40] The effect is included in Fig. 3 although no detailed discussion is provided in this article.…”

Technical Letter Report on the Cracking of Irradiated Stainless Steels in Low-Corrosion-Potential Environments

“…Figure 46 shows the estimated fracture toughness values obtained in this study along with the unirradiated fracture toughness values of 304 and 316 SSs reported by Mills 30,31 and some previous results from Halden-I and -II irradiated specimens. 15,26 Despite a difference in test temperature, a decreasing trend can be seen with increasing dose for this group of data. Among the samples tested in the current study, SA 304L shows the best fracture resistance, and the measured J at 0.2-mm offset line was about 95 kJ/m 2 .…”

“…This strong effect of PPU was different from that observed at lower doses and in BWR NWC (a high-corrosionpotential environment). 15 Periodic partial unloading seemed not to have a substantial effect on the measured CGRs in high-purity water with 700-800 ppb DO for irradiated SSs. 15 The different sensitivity to partial unloading suggests a combined effect of SCC and low-cycle fatigue on the crack growth response of irradiated SSs in the low-corrosion-potential environments.…”

“…13,14 For irradiated SSs, a similar correlation between the CGR and ECP has also been observed. [15][16][17] While the cracking behavior of irradiated SSs in BWR normal water chemistry (NWC) has been investigated in several programs in recent years, a systematic analysis of the cracking of irradiated materials in the low corrosion potential regime is still lacking. The CGR database of irradiated SSs is limited for BWR HWC or PWR environments at present.…”