Intergranular stress-assisted corrosion cracking of austenitic alloys in water-cooled nuclear reactors

Cheng, C. F.

doi:10.1016/0022-3115(75)90175-0

Cited by 51 publications

(10 citation statements)

References 9 publications

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“…[1][2][3][4][5][6] Failures of the fuel cladding, control rod cladding, and core baffle bolts in PWRs have revealed intergranular (IG) fracture similar to the failure observed in the core internals of boiling water reactors (BWRs). 3,4,6 However, because of the low corrosion potential, the typical "threshold" neutron dose for IASCC in PWRs is about one order of magnitude higher than that in BWRs with normal water chemistry. 5 Also, chromium depletion at grain boundaries is considered as less critical for IASCC sensitivity in PWRs than in BWRs.…”

Section: Introductionmentioning

confidence: 99%

Cracking behavior and microstructure of austenitic stainless steels and alloy 690 irradiated in BOR-60 reactor, phase I.

Chen¹,

Chopra²,

Soppet³

et al. 2010

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Cracking behavior of stainless steels specimens irradiated in the BOR-60 at about 320°C is studied. The primary objective of this research is to improve the mechanistic understanding of irradiation-assisted stress corrosion cracking (IASCC) of core internal components under conditions relevant to pressurized water reactors. The current report covers several baseline tests in air, a comparison study in highdissolved-oxygen environment, and TEM characterization of irradiation defect structure.Slow strain rate tensile (SSRT) tests were conducted in air and in high-dissolved-oxygen (DO) water with selected 5-and 10-dpa specimens. The results in high-DO water were compared with those from earlier tests with identical materials irradiated in the Halden reactor to a similar dose. The SSRT tests produced similar results among different materials irradiated in the Halden and BOR-60 reactors. However, the post-irradiation strength for the BOR-60 specimens was consistently lower than that of the corresponding Halden specimens. The elongation of the BOR-60 specimens was also greater than that of their Halden specimens. Intergranular cracking in high-DO water was consistent for most of the tested materials in the Halden and BOR-60 irradiations. Nonetheless, the BOR-60 irradiation was somewhat less effective in stimulating IG fracture among the tested materials.Microstructural characterization was also carried out using transmission electron microscopy on selected BOR-60 specimens irradiated to 25 dpa. No voids were observed in irradiated austenitic stainless steels and cast stainless steels, while a few voids were found in base and grain-boundaryengineered Alloy 690. All the irradiated microstructures were dominated by a high density of Frank loops, which varied in mean size and density for different alloys.

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Section: Introductionmentioning

confidence: 99%

Cracking behavior and microstructure of austenitic stainless steels and alloy 690 irradiated in BOR-60 reactor, phase I.

Chen¹,

Chopra²,

Soppet³

et al. 2010

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“…IASCC is clearly not confi ned to a particular reactor design, material, component or water chemistry. For example, stainless steel fuel cladding failures were reported years ago in commercial PWRs and in PWR test reactors (Andresen et al, 1990;Brown Jr et al, 1967;Cheng, 1964Cheng, , 1970Cheng, , 1975Garzarolli, Rubel, & Steinberg, 1984;Hanninen & AhoMantila, 1988;Multer, 1975;Pasupathi & Klingensmith, 1981;Schaffer, 1962;Storrer & Locke, 1970). At the West Milton PWR test loop, intergranular failure of vacuum annealed type 304 stainless steel fuel cladding was observed (Cheng, 1970) in 316 ° C ammoniated water (pH 10) when the cladding was stressed above yield.…”

Section: Plant Data On Iasccmentioning

confidence: 94%

“…Subsequent observation made this clearer e.g., by showing a similar effect of the water purity (as measured solution conductivity) on unirradiated and irradiated stainless steel (Figure 4 ). The initial reports of IASCC occurred in the early 1960s (Andresen & Ford, 1995;Andresen, Ford, Murphy, & Perks, 1990a;Armijo, Low, & Wolff, 1965;Brown, Storhok, & Gates, 1967;Bruemmer et al, 1999;Cheng, 1964Cheng, , 1970Cheng, , 1975Cowan & Gordon, 1977;Duncan et al, 1965;Hanninen & Aho-Mantila, 1988;Jacobs & Wozadlo, 1985;Multer, 1975;Pashos et al, 1964;Pasupathi & Klingensmith, 1981;Schaffer, 1962;Scott, 1994;Storrer & Locke, 1970;Was & Andresen, 1992a;2007) and involved intergranular cracking of stainless steel fuel cladding initiating from the water side. Post-irradiation mechanical tests, by comparison, were performed in inert environments at various strain rates and temperatures and exhibited mostly ductile, transgranular cracking.…”

Section: Plant Data On Iasccmentioning

confidence: 95%

Irradiation-assisted stress corrosion cracking

Was¹,

Ashida²,

Andresen³

2011

Corrosion Reviews

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Irradiation-assisted stress corrosion cracking (IASCC) refers to the process(es) by which irradiation enhances the inherent susceptibility of alloys to stress corrosion cracking (SCC). This article begins with an introduction to the IASCC problem experienced in light water reactors, followed by a summary of existing plant and laboratory data that reveal its dependence on irradiation, environment and alloy parameters. The specifi c effects of irradiation on IASCC are classifi ed into two categories: the effects on water chemistry and the effects on microstructure. Water chemistry effects include radiolysis and its effects on corrosion potential, and effects of corrosion potential on IASCC. Microstructure effects include radiationinduced segregation (RIS), irradiated microstructure, swelling and creep, and hydrogen and helium generation. Leading mechanisms proposed to explain the role of radiation in the SCC process are: radiolysis and crack tip strain rate, grain boundary chromium depletion, irradiation hardening, localized deformation and RIS of minor elements. As nuclear power plants operate for longer, an increased incidence of SCC can be expected unless active mitigation steps are taken. Drawing on the accumulated understanding over the past four decades of the key processes believed to control IASCC, a set of attributes of an ideal, IASCC-resistant alloy are identifi ed as a fi rst step.

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“…Service failures of fuel cladding, internal structures, and fasteners have been discovered in both boiling water reactors (BWRs) and pressurized water reactors (PWRs). [3][4][5][6][7] Since repair or replacement of structural components at the reactor core regions is extremely difficult and expensive, failures of major internal components could seriously impair the safe and economic operation of LWRs. As nuclear power plants age and accumulated neutron fluence increases in recent years, IASCC has become a progressively more important issue for license renewal and aging management worldwide.…”

Section: Executive Summarymentioning

confidence: 99%

“…2 The CGR at the end of the test period is reported. 3 Lower the load to minimize crack growth prior to changing test conditions. 4 Inadequate load shed resulting in a rising K condition.…”

Section: Cgr and J-r Curve Testsmentioning

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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Intergranular stress-assisted corrosion cracking of austenitic alloys in water-cooled nuclear reactors

Cited by 51 publications

References 9 publications

Cracking behavior and microstructure of austenitic stainless steels and alloy 690 irradiated in BOR-60 reactor, phase I.

Cracking behavior and microstructure of austenitic stainless steels and alloy 690 irradiated in BOR-60 reactor, phase I.

Irradiation-assisted stress corrosion cracking

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

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