Relationship between evolution of mechanical properties of various cast duplex stainless steels and metallurgical and aging parameters: outline of current EDF programmes

Bonnet, Stéphanie; Bourgoin, J.; Champredonde, J.; Guttmann, D.; Guttmann, M.

doi:10.1179/026708390790191161

Cited by 33 publications

(37 citation statements)

References 2 publications

Supporting

Mentioning

Contrasting

Order By: Relevance

“…Formation of Cr-rich ∃% phase in the ferrite is the primary mechanism for thermal embrittlement of cast austenitic SSs; [28][29][30][31][32][33][34][35][36] thermal aging has little or no effect on the austenite phase. Embrittlement of ferrite phase from neutron irradiation occurs at lower fluences than does embrittlement of the austenite phase.…”

Section: Synergistic Effect Of Thermal and Neutron Irradiationmentioning

confidence: 99%

“…27 Another issue that has been a concern for reactor core internal components is the possibility of a synergistic interaction between irradiation and thermal embrittlement of cast austenitic SSs and SS weld metals. [28][29][30][31][32] Although wrought SSs are typically completely austenitic, welded and cast SSs have a duplex microstructure consisting of austenite and ferrite phases. The ferrite phase increases the tensile strength and improves resistance to SCC, but it is susceptible to thermal embrittlement after extended service at reactor operating temperatures.…”

Section: Introductionmentioning

confidence: 99%

See 1 more Smart Citation

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

Chopra¹,

Shack²

2008

View full text Add to dashboard Cite

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 ...

show abstract

Section: Synergistic Effect Of Thermal and Neutron Irradiationmentioning

confidence: 99%

Section: Introductionmentioning

confidence: 99%

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

Chopra¹,

Shack²

2008

View full text Add to dashboard Cite

show abstract

“…Activation energies of thermal embrittlement range from 65 to 230 kJ/mole (15 to 55 kcal/mole). [4][5][6]11,[13][14][15]22 These values are well below the 202 kJ/mole (48 kcal/mole) value associated with Cr bulk diffusion in the Fe-28Cr alloy. 23 Small changes in the constituent elements of the material can cause the kinetics of thermal embrittlement to vary significantly.…”

Section: Mechanism Of Thermal Embrittlementmentioning

confidence: 99%

“…13,14,22 Microstructural characterization and annealing studies on thermally aged cast stainless steel show that strengthening of ferrite is caused primarily by spinodal decomposition of ferrite to form the Cr-rich α' phase. 4,9,10 Consequently, the kinetics of thermal embrittlement should be controlled by the amplitude and frequency of Cr fluctuations produced by spinodal decomposition, i.e., by the size and spacing of the α' phase.…”

Section: Mechanism Of Thermal Embrittlementmentioning

confidence: 99%

“…Investigations at Argonne National Laboratory (ANL) [1][2][3][4][5][6][7][8][9][10] and elsewhere [11][12][13][14][15][16] have shown that thermal embrittlement of cast stainless steel components occurs during the reactor design lifetime of 40 yr. Various grades and heats of cast stainless steel exhibit varying degrees of thermal embrittlement.…”

Section: Introductionmentioning

confidence: 99%

See 1 more Smart Citation

Tensile-property characterization of thermally aged cast stainless steels.

Michaud¹,

Toben²,

Soppet³

et al. 1994

View full text Add to dashboard Cite

The effect of thermal aging on tensile properties of cast stainless steels during service in light water reactors has been evaluated. Tensile data for several experimental and commercial heats of cast stainless steels are presented. Thermal aging increases the tensile strength of these steels. The high-C Mo-bearing CF-8M steels are more susceptible to thermal aging than the Mo-free CF-3 or CF-8 steels. A procedure and correlations are presented for predicting the change in tensile flow and yield stresses and engineering stress-vs.-strain curve of cast stainless steel as a function of time and temperature of service. The tensile properties of aged cast stainless steel are estimated from known material information, i.e., chemical composition and the initial tensile strength of the steel. The correlations described in this report may be used for assessing thermal embrittlement of cast stainless steel components. Executive SummaryCast duplex stainless steels used in light water reactor (LWR) systems for primary pressure-boundary components are susceptible to thermal embrittlement at reactor operating temperatures. Thermal aging of cast stainless steels at these temperatures causes an increase in hardness and tensile strength and a decrease in ductility, impact strength, and fracture toughness of the material and the Charpy transition curve shifts to higher temperatures. Investigations at Argonne National Laboratory have shown that thermal embrittlement of cast stainless steel components may occur within the reactor design lifetime of 40 yr. Various grades and heats of cast stainless steel exhibit varying degrees of thermal embrittlement. In general, the low-C CF-3 steels are the most resistant t o thermal embrittlement, and the Mobearing, high-C CF-8M steels are the least resistant. An assessment of mechanical-property degradation due to thermal embrittlement is therefore required to evaluate the performance of cast stainless steel components during prolonged exposure to service temperatures, because rupture of the primary pressure boundary could lead to a loss-of-coolant accident and possible exposure of the public to radiation.This report presents tensile-property data on several heats of cast stainless steels aged up to 58,000 h at temperatures between 290 and 450°C (554 and 752°F). The tensile data are analyzed to establish the effects of thermal aging on tensile strength and engineering stress-strain behavior (represented by the Ramberg-Osgood equation) of cast stainless steels. A procedure and correlations are presented for predicting the change in tensile flow and yield stress, and in the engineering stress-vs.-strain curve of cast stainless steel components due to thermal aging during service in LWRs. The tensile properties of aged cast stainless steel are estimated from information that is readily available from certified material test records for the component, i.e., chemical composition and the initial tensile strength of the unaged material. The correlations described in this report may be used ...

show abstract

Phase Transformation and Microstructure

Padilha

Plaut

2013

Duplex Stainless Steels

View full text Add to dashboard Cite

Relationship between evolution of mechanical properties of various cast duplex stainless steels and metallurgical and aging parameters: outline of current EDF programmes

Cited by 33 publications

References 2 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.

Tensile-property characterization of thermally aged cast stainless steels.

Phase Transformation and Microstructure

Contact Info

Product

Resources

About