Stress Corrosion Cracking Susceptibility of Irradiated Type 304 Stainless Steels

Chung, H.M.; Ruther, W.E.; Sanecki, J.E.; Hins, A.G.; Kassner, T.F.

doi:10.1520/stp23976s

Cited by 8 publications

(9 citation statements)

References 2 publications

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“…A description of the hot-cell SSRT tests was given in a previous report . 42 Cylindrical SSRT specimens 89-mm long were sectioned from top-, middle-, and bottom-axial positions of BWR neutron-absorber rods listed in Table 10. Boron carbide was removed by drilling.…”

Section: Methodsmentioning

confidence: 99%

“…3 ppm DO and an ECP of -60-150 mV SHE. 42 20-23, 1994. water chemistry were investigated for a wider range of DO and ECP. The objective was twofold: (1) to better define the effects of water chemistry on susceptibility of the HP and CP Type 304 SS BWR components to IG cracking, thereby providing an independent confirmation of the protection potential: and (2) to obtain insight into the origin of heatto-heat variations in susceptibility to IASCC.…”

Section: Introductionmentioning

confidence: 99%

See 1 more Smart Citation

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

Chopra¹,

Chung²,

Gavenda³

1995

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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 Lbrbon 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 q p e 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 S S s and Alloy 600 to investigate EAC in simulated boiling water reactor (BWR) and pressurized water reactor environments at 289°C. The data were compared with predictions from crack growth correlations developed at ANL for S S s in water and from rates in air from Section X I 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. . 6. 7.8. 9.10. 11.12. 4. 5.6. 7.8 . 9.10. 11.12. 13.14. Executive Summarv Fatigue of Ferritic Piping and Pressure Vessel SteelsPlain carbon and low-alloy steels 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 A106-Gr B carbon steel and A533-Gr B and A302-Gr B low-alloy steels in water and in air at 288°C to establish the effects of material and loading variables on fatigue life. The results indicate that in water with high dissolved-oxygen (DO) content, a minimum strain is required for environmentally assisted decrease in fatigue life of these steels, and a slow strain rate applied during the tensile-loading cycle decreases fatigue life more than when applied during the compressive-loading cycle. During the present reporting period, several exploratory tests were conducted on A106-Gr B and A533-Gr B steels in high-DO water in which the slow strain rate was applied during only a portion of the tensile-loading cycle to check whether each portion of the tensile cycle is equally effective in decreasing fatigue life in high-DO water. The results indicate that slow strain rates applied during any portion of the tensile-loading cycle above the threshold strain are equally damaging.The effects of various material and loading variables, e.g., steel type, strain rate, DO, strain range, loading waveform, and surface morphology, on fatigue life of carbon and low-alloy steels in our tests are summarized in this report. Environmentally Assisted Cracking of Alloy 600 and Wrought Stainless Steels in Simulated LWR WaterFr...

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Section: Methodsmentioning

confidence: 99%

Section: Introductionmentioning

confidence: 99%

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

Chopra¹,

Chung²,

Gavenda³

1995

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“…Fluorine analyses by AES were conducted on specimens from BWR-irradiated components: the experimental methods were similar to those reported earlier. [55][56][57] The specimens were obtained from HP-and CP-grade '&pe 304 SS neutron-absorber tubes and a control-blade sheath discharged from several BWRS after irradiation up to 2.6 x 1021 ncm-2 (E > 1 MeV). Expanding-mandrel tube specimens of irradiated Type 348 SS were also analyzed by AES.…”

Section: Irradiation-assisted Scc Of Austenitic Stainless Steelsmentioning

confidence: 99%

Environmentally assisted cracking in light water reactors.

Chopra¹,

Chung²,

Clark³

et al. 2007

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The following documents In the NUREG series are available for purchase from the Government Printing Office:formal NRC staff and contractor reports, NRC-sponsored conference proceedings, international agreement reports, grantee reports, and NRC booklets and brochures. Also available a r e regulatory guides, NRC regula- Fatigue tests were conducted on ferritic steels in water that contained various concentrations of dissolved oxygen (DO) to determine whether a slow strain rate applied during different portions of a tensileloading cycle are equally effective in decreasing fatigue life. Crack-growth-rate tests were conducted on compact-tension specimens from several heats of Alloys 600 and 690 in simulated LWR environments. Effects of fluoride-ion contamination on susceptibility to intergranular cracking of high-and commercial-purity Type 304 SS specimens from controlblade absorber tubes irradiated in boiling water reactors were determined in slow-strain-ratetensile tests at 288OC. Microchemical changes in the specimens were studied by Auger electron spectroscopy and scanning electron microscopy to determine whether trace impurity elements may contribute to IASCC of these materials.

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“…The IASCCresistant or -susceptible behavior at ≈ 3 dpa is largely determined by the bulk concentrations of S and C. Similar analysis was performed for data reported in the literature. [17][18][19][20][25][26][27][28][29][30][31][32][33][34][35][36][37][38] Virtually all literature data have been reported after 1987. The two-dimensional map of bulk S and C contents in Figs.…”

Section: Representation Of Iascc-resistant or -Susceptible Behavior Omentioning

confidence: 99%

“…One of the most important factors that seem to support the Cr-depletion mechanism is the observation that the dependence on water chemistry (i.e., oxidizing potential) of intergranular stress corrosion cracking (IGSCC) of nonirradiated thermally sensitized material and of IASCC of BWR-irradiated solution-annealed material is similar. 14,17,18 Many investigators have also suggested that radiation-induced segregation of ASTM-specified impurities such as Si and P and other minor impurities not specified in the ASTM specification [21][22][23][24][25][26][27][28][29][30][31][32][33][34][35][36] has a role in the IASCC process. However, the exact mechanism of IASCC still remains unknown.…”

Section: Introductionmentioning

confidence: 99%

Environmentally assisted cracking in light water reactors : Annual Report, January-December 2003.

Alexandreanu¹,

Chopra²,

Chung³

et al. 2005

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This report summarizes work performed by Argonne National Laboratory on fatigue and environmentally assisted cracking (EAC) in light water reactors (LWRs) from January to December 2003. Topics that have been investigated include: (a) environmental effects on fatigue crack initiation in carbon and low-alloy steels and austenitic stainless steels (SSs), (b) irradiation-assisted stress corrosion cracking (IASCC) of austenitic SSs in boiling water reactors (BWRs), (c) evaluation of causes and mechanisms of irradiation-assisted cracking of austenitic SS in pressurized water reactors (PWRs), and (d) cracking in Ni alloys and welds.Fatigue tests have been conducted on two heats of Type 304 stainless steel (SS) under various material conditions to determine the effect of heat treatment on fatigue crack initiation in these steels in air and LWR environments. Heat treatment has little or no effect on the fatigue life in air and low dissolved oxygen (DO) environment, whereas in a high-DO environment, fatigue life is lower for sensitized SSs.Crack growth rate (CGR) data were obtained on Type 304L SS (Heat C3) irradiated to 0.3 x 10 21 n/cm 2 , nonirradiated Type 304 L SS weld heat affected zone (HAZ) specimens from the Grand Gulf (GG) reactor core shroud, and a Type 304 SS laboratory-prepared weld. The irradiated specimen of Heat C3 showed very little enhancement of CGRs in high-DO water. The results for the weld HAZ material indicate that under predominantly mechanical fatigue conditions, the CGRs for the GG Type 304L weld HAZ are lower than those for shielded metal arc (SMA) weld HAZ prepared in the laboratory with Type 304 SS.Slow-strain-rate tensile (SSRT) tests have been completed in high-purity 289°C water on steels irradiated to ≈3 dpa. The bulk S content correlated well with the susceptibility to intergranular stress corrosion cracking (IGSCC) in 289°C water. The irradiation-assisted stress corrosion cracking (IASCC) susceptibility of SSs that contain >0.003 wt.% S increased drastically. These results and a review of other data in the literature indicate that IASCC in 289°C water is dominated by a crack-tip grain-boundary process that involves S. The IASCC-resistant or -susceptible behavior of austenitic SSs in BWR-like oxidizing environment is described in terms of a two-dimensional map of bulk S and C contents of the steels.Crack growth tests were completed on a Alloy 600 round robin specimen and a Alloy 182 weld specimen in simulated PWR water at 320°C. Under cyclic loading, the CGRs for the weld specimen were a factor of ≈5 higher than those for Alloy 600 under the same loading conditions; little or no environmental enhancement was observed. The CGRs obtained with a trapezoidal waveform (i.e., a constant load with periodic unload/reload) were comparable to the average behavior of Alloy 600 in a PWR environment. The cyclic CGRs for the Alloy 600 round-robin specimen were a factor of ≈2 lower than those predicted in air. However, the crack front was U-shaped, indicating that the growth rates were significa...

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Stress Corrosion Cracking Susceptibility of Irradiated Type 304 Stainless Steels

Cited by 8 publications

References 2 publications

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

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

Environmentally assisted cracking in light water reactors.

Environmentally assisted cracking in light water reactors : Annual Report, January-December 2003.

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