Crack growth rates of irradiated austenitic stainless steel weld heat affected zone in BWR environments.

Chopra, O.K.; Alexandreanu, B.; Gruber, E.E.; Daum, Robert S.; Shack, W. J.

doi:10.2172/925223

Cited by 14 publications

(33 citation statements)

References 8 publications

(8 reference statements)

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“…The cyclic CGRs from the present study and those obtained earlier at ANL 27,54 are compiled in Table 15. The tests were conducted with either a triangular or slow/fast sawtooth waveform.…”

Section: Cgrs Under Continuous Cyclingmentioning

confidence: 97%

“…53 For the few materials that were tested in air and water environments, the experimental tensile stress was 10-20% higher in air than in water. Table 3 lists the tensile properties of nonirradiated Type 304L SS from the GG core shroud shell, Heat 10285 of Type 304 SS in the mill-annealed condition and after sensitization at 600°C for 10.5 h, 54 and the thermally aged cast CF-8M SS. 28 For these steels, the tensile properties of the irradiated materials have not been measured and were therefore estimated.…”

Section: Alloys and Specimen Preparationmentioning

confidence: 99%

“…The experimental CGRs for nonirradiated SS weld HAZ materials in high-DO water 54 and those predicted in air for the same loading conditions are plotted in Fig. 67; the loading conditions for the data points shown with a "+" did not satisfy the K/size criterion of ASTM E-647.…”

Section: Simulated Bwr Environmentmentioning

confidence: 99%

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

Chopra¹,

Shack²

2008

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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 ...

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“…The cyclic CGRs from the present study and those obtained earlier at ANL 27,54 are compiled in Table 15. The tests were conducted with either a triangular or slow/fast sawtooth waveform.…”

Section: Cgrs Under Continuous Cyclingmentioning

confidence: 97%

Section: Alloys and Specimen Preparationmentioning

confidence: 99%

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

Chopra¹,

Shack²

2008

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“…While in most cases HWC has a beneficial effect, several studies have found that the crack propagation behavior fails to respond to a decline in corrosion potential in some high-dose SS specimens. [18][19][20] These observations suggest that the mitigation of HWC may only be effective within a certain dose range. Even in cases where the mitigation effect of HWC is observed, the impact of neutron irradiation on the dependence of CGR on corrosion potential is not well understood.…”

Section: Bwrmentioning

confidence: 99%

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

Chen¹,

Alexandreanu²,

Natesan³

2013

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Irradiation-assisted stress corrosion cracking (IASCC) is a major degradation mechanism for reactor internal components of austenitic stainless steels (SSs) in light water reactors. Crack growth rate (CGR) and fracture toughness J-R curve tests were performed on four irradiated SS specimens in high-purity water with low dissolved oxygen and in a simulated pressurized water reactor (PWR) environment. The samples had been previously irradiated from ~5 to 8 dpa in the BOR-60 reactor. The materials were cold-worked (CW) 316 and solution-annealed (SA) 304L SSs, which are commonly used in PWR core internals. Cyclic and constant-load CGRs were obtained from these samples to assess their cracking behaviors in low-corrosion-potential environments. A Ti-stabilized specimen, CW 316-Ti SS, was found to be highly vulnerable to IASCC, and its cracking susceptibility increased with neutron dose between ~5 and 8 dpa. The CW 316 and SA 304L SSs showed moderate cracking susceptibilities in the low-corrosionpotential environments up to ~7 dpa. Periodic partial unloading (PPU) had a significant impact on the crack growth behavior of all irradiated specimens, and the measured CGRs were more than one order of magnitude higher in the test periods with than without PPU. A "stair-step" crack growth behavior was also observed for all irradiated specimens during the test periods with PPU, suggesting a strong effect of dynamic loading on IASCC. The SA 304L SS showed the best fracture resistance, and the CW 316 SS was the most brittle among the tested specimens. The fracture toughness J values of the two CW 316-Ti SS specimens were nearly identical at ~5 and ~8 dpa.

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“…Durante o processo de soldagem pode ocorrer o aparecimento de alguns problemas como o crescimento de grãos, aparecimento de estruturas frágeis e fissurações, dentre outros [1][2][3][4][5]. Tais problemas metalúrgicos são freqüentemente associados à zona afetada pelo calor (ZAC) [6][7][8][9] de dimensões reduzidas.…”

Section: Introductionunclassified

Metodologia de baixo custo para levantamento de diagramas CCT em soldagem

Vilarinho

Araújo

2010

Soldag. insp.

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Crack growth rates of irradiated austenitic stainless steel weld heat affected zone in BWR environments.

Cited by 14 publications

References 8 publications

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

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

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

Metodologia de baixo custo para levantamento de diagramas CCT em soldagem

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