Fracture Toughness of Irradiated AISI 304 and 316L Stainless Steels

Dufresne, J.; Henry, B.; Larsson, Henrik

doi:10.1520/stp38185s

Cited by 19 publications

(5 citation statements)

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“…1. [10][11][12][13][14][15][16][17][18] The effects of irradiation may be divided into three regimes: little or no loss of toughness below a threshold exposure of ª1 dpa, substantial decrease in toughness at exposures of 1-10 dpa, and no further reduction in toughness above a saturation exposure of 10 dpa. The effect is largest in high-toughness steels.…”

Section: Introductionmentioning

confidence: 99%

Fracture toughness and crack growth rates of irradiated austenitic stainless steels.

Chopra¹,

Gruber²,

Shack³

2003

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Austenitic stainless steels (SSs) are used extensively as structural alloys in the internal components of reactor pressure vessels because of their superior fracture toughness properties. However, exposure to high levels of neutron irradiation for extended periods leads to significant reduction in the fracture resistance of these steels. Experimental data are presented on fracture toughness and crack growth rates (CGRs) of austenitic SSs irradiated to fluence levels up to 2.0 x 10 21 n/cm 2 (E > 1 MeV) (ª3.0 dpa) at ª288°C. Crack growth tests were conducted under cycling loading and long hold time trapezoidal loading in simulated boiling water reactor (BWR) environments, and fracture toughness tests were conducted in air. Neutron irradiation at 288°C decreases the fracture toughness of the steels; the data from commercial heats fall within the scatter band for the data obtained at higher temperatures. In addition, the results indicate significant enhancement of CGRs of the irradiated steels in normal water chemistry BWR environment; the CGRs for irradiated steels are a factor of ª5 higher than the disposition curve proposed for sensitized austenitic SSs. The rates decreased by more than an order of magnitude in low-dissolved-oxygen BWR environment.

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Section: Introductionmentioning

confidence: 99%

Fracture toughness and crack growth rates of irradiated austenitic stainless steels.

Chopra¹,

Gruber²,

Shack³

2003

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“…At around 300°C, fracture toughness for nonirradiated austenitic SSs is reported to be in the range of 200-600 kJ/m 2 . [65][66][67] Weld process and neutron irradiation both result in significant reductions in fracture toughness by introducing new embrittlement mechanisms. In the present study, fracture toughness measured in NWC BWR water was found to be similar between the SA and SMA HAZ specimens under the as-welded condition (Fig.…”

Section: Fracture Toughness Of Weld Haz Specimensmentioning

confidence: 99%

Irradiation-Assisted Stress Corrosion Cracking of Austenitic Stainless Steels in BWR Environments

Chen

Chopra

Gruber

et al. 2010

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The internal components of light water reactors are exposed to high-energy neutron irradiation and high-temperature reactor coolant. The exposure to neutron irradiation increases the susceptibility of austenitic stainless steels (SSs) to stress corrosion cracking (SCC) because of the elevated corrosion potential of the reactor coolant and the introduction of new embrittlement mechanisms through radiation damage. Various nonsensitized SSs and nickel alloys have been found to be prone to intergranular cracking after extended neutron exposure. Such cracks have been seen in a number of internal components in boiling water reactors (BWRs). The elevated susceptibility to SCC in irradiated materials, commonly referred to as irradiation-assisted stress corrosion cracking (IASCC), is a complex phenomenon that involves simultaneous actions of irradiation, stress, and corrosion. In recent years, as nuclear power plants have aged and irradiation dose increased, IASCC has become an increasingly important issue. Post-irradiation crack growth rate and fracture toughness tests have been performed to provide data and technical support for the NRC to address various issues related to aging degradation of reactor-core internal structures and components. This report summarizes the results of the last group of tests on compact tension specimens from the Halden-II irradiation. The IASCC susceptibility of austenitic SSs and heat-affected-zone (HAZ) materials sectioned from submerged arc and shielded metal arc welds was evaluated by conducting crack growth rate and fracture toughness tests in a simulated BWR environment. The fracture and cracking behavior of HAZ materials, thermally sensitized SSs and grain-boundary engineered SSs was investigated at several doses (3 dpa). These latest results were combined with previous results from Halden-I and II irradiations to analyze the effects of neutron dose, water chemistry, alloy compositions, and welding and processing conditions on IASCC. The effect of neutron irradiation on the fracture toughness of austenitic SSs was also evaluated at dose levels relevant to BWR internals. Paperwork Reduction Act Statement This 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 Notification The 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. vi sensitized at 600 o C for 10.5 and 24 hours, the degree of sensitization has a negligible effect on IASCC resistance. Also, the GBE treatment (the fraction of coincident-site-lattice boundaries ~ 60%) did not improve IASCC susceptibility in the test materials and environment. Conversely, increasing the neutron fluence significantly decreased both the IASCC resistance and the fracture toughness. The research described in this report provides relevant test da...

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“…21. [81][82][83][84][85][86][87][88][89] The effects of irradiation may be divided into three regimes: little or no loss of toughness below a threshold exposure of =1 dpa, substantial decrease in toughness at exposures of 1-10 dpa, and no effect on toughness above a saturation exposure of 10 dpa. The effect is largest in high-toughness steels.…”

Section: Fracture Toughness J-r Test Of Austenitic Stainless Steels Imentioning

confidence: 99%

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

Chopra¹,

Chung²,

Gruber³

et al. 1999

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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 (

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Fracture Toughness of Irradiated AISI 304 and 316L Stainless Steels

Cited by 19 publications

References 0 publications

Fracture toughness and crack growth rates of irradiated austenitic stainless steels.

Fracture toughness and crack growth rates of irradiated austenitic stainless steels.

Irradiation-Assisted Stress Corrosion Cracking of Austenitic Stainless Steels in BWR Environments

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

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