Analysis of tensile deformation and failure in austenitic stainless steels: Part I - Temperature dependence

This paper describes the temperature dependence of deformation and failure behaviors in the austenitic stainless steels (annealed 304, 316, 316LN, and 20% cold-worked 316LN) in terms of equivalent true stress-true strain curves. The true stress-true strain curves up to the final fracture were calculated from tensile test data obtained at −150 to 450 °C using an iterative finite element method. Analysis was largely focused on the necking and fracture: key parameters such as the strain hardening rate, equivalent fracture stress, fracture strain, and tensile fracture energy were evaluated, and their temperature dependencies were investigated. It was shown that a significantly high strain hardening rate was retained during unstable deformation although overall strain hardening rate beyond the onset of necking was lower than that of the uniform deformation. The fracture stress and energy decreased with temperature up to 200 °C and were nearly saturated as the temperature came close to the maximum test temperature 450 °C. The fracture strain had a maximum at −50 to 20 °C before decreasing with temperature. It was explained that these temperature dependencies of fracture properties were associated with a change in the dominant strain hardening mechanism with test temperature. Also, it was seen that the pre-straining of material has little effect on the strain hardening rate during necking deformation and on fracture properties.     

Irradiation creep and interrelation with swelling in austenitic stainless steels

The available experimental data on irradiation-induced creep in austenitic stainless steels are summarized and the existing theories reviewed. Attention is paid to the influence of material composition and pretreatments on irradiation creep. In particular the stress, flux, fluence and temperature dependencies are reported and possible correlations of irradiation creep with the microstructural evolution, the swelling behaviour and the precipitation kinetics of the materials are outlined. The consequences of stress effects connected with swelling for the irradiation-creep behaviour, especially the stress-dependence, are discussed.     

Impact of localized deformation on IASCC in austenitic stainless steels

Localized deformation has been identified as a potential primary contributor to IASCC. Seven austenitic alloys were irradiated to 1 and 5 dpa at 360 °C using 2-3.2 MeV protons and were tested both in simulated BWR environment and in argon. Cracking susceptibility was evaluated at both 1% and 3% strain intervals using crack length per unit area. Stacking fault energy (SFE), hardness, radiation-induced segregation (RIS) and localized deformation were characterized and their correlations with cracking were evaluated using a proposed term, correlation strength. Both SFE and hardness contributed to cracking but neither was the dominant factor. RIS did not play an important role in this study. The correlation strength of localized deformation with IASCC was found to be significantly higher than for others parameters, implying that localized deformation is the most important factor in IASCC. Although not well understood, localized deformation may promote cracking through intensive interaction of dislocations in slip channels with grain boundaries.     

Irradiation damage in 304 and 316 stainless steels: experimental investigation and modeling. Part II: Irradiation induced hardening

The consequences of irradiation damage in austenitic stainless steels on their mechanical properties, namely the yield stress, are investigated both experimentally and theoretically. The observed hardening is correlated with the quantitative characteristics of irradiation defects population. A simple model allowing for the defaulting of Frank loops under stress predicts the hardening and its saturation at large doses.     

The role of irradiated microstructure in the localized deformation of austenitic stainless steels

Localized deformation has emerged as a potential factor in irradiation-assisted stress corrosion cracking of austenitic stainless steels in LWR environments and the irradiated microstructure may be a critical factor in controlling the degree of localized deformation. Seven austenitic alloys with various compositions were irradiated using 2-3 MeV protons to doses of 1 and 5 dpa at 360 °C. The irradiated microstructure consisting of dislocation loops and voids was characterized using transmission electron microscopy. The degree of localized deformation was characterized using atomic force microscopy on the deformed samples after conducting constant extension rate tension tests to 1% and 3% strain in argon. Localized deformation was found to be dependent on the irradiated microstructure and to correlate with hardening originating from dislocation loops. Dislocation loops enhance the formation of dislocation channels and localize deformation into existing channels. On the contrast, voids mitigate the degree of localized deformation. The degree of localized deformation decreases with SFE with the exception of alloy B. Localized deformation was found to have similar dependence on SFE as loop density suggesting that SFE affects localized deformation by altering irradiated microstructure.     

Precipitation and void-swelling in nickel-manganese austenitic stainless steels

The precipitation and void-swelling characteristics of austenitic stainless steels in which nickel is partially replaced by manganese have been investigated. Alloy compositions were chosen on the basis of manganese being half as effective as nickel in stabilizing austenite, and steels with “nickel equivalent” contents of 25–37% were examined. The steels were irradiated with 46 MeV Ni⁶⁺ ions to 60 dpa at 625°C and also aged for 1000 h at 600°C. The high-Mn alloys (20–30% Mn) were very susceptible to the formation of intermetallic phases during thermal ageing but less so in the shorter-duration irradiation experiment. Irradiation promoted the formation of Ni- and Si-rich phases—the suicide G phase (in which Mn can replace Ti) and in one instance M₆C. The Cr-rich carbide M₂₃C₆ formed in both the aged and irradiated steels. Among the high-Mn alloys, void-swelling decreased with increasing Ni and (Ni+Mn) contents, although a 25Ni-1Mn steel showed no swelling at 625°C.     

Fracture behavior of austenitic stainless steels irradiated in PWR

Fracture behavior of cold-worked 316 stainless steels irradiated up to 73 dpa in a pressurized water reactor was investigated by impact testing at −196, 30 and 150 °C, and by conventional tensile and slow tensile testing at 30 and 320 °C. In impact tests, brittle IG mode was dominant at −196 °C at doses higher than 11 dpa accompanying significant decrease in absorbed energy. The mixed IG mode, which was characterized by isolated grain facets in ductile dimples, appeared at 30 and 150 °C whereas the fracture occurred macroscopically in a ductile manner. The sensitivity to IG or mixed IG mode was more pronounced for higher dose and lower test temperature. In uniaxial tensile tests, IG mode at a slow strain rate appeared only at 320 °C whereas mixed IG mode appeared at both 30 and 320 °C at a fast strain rate. A compilation of the results and literature data suggested that IG fracture exists in two different conditions, low-temperature high-strain-rate (LTHR) and high-temperature low-strain-rate (HTLR) conditions. These two conditions for IG fracture likely correspond to two different deformation modes, twining and channeling.     

TEM observations and finite element modelling of channel deformation in pre-irradiated austenitic stainless steels - Interactions with free surfaces and grain boundaries

Transmission electron microscopy (TEM) observations show that dislocation channel deformation occurs in pre-irradiated austenitic stainless steels, even at low stress levels (∼175 MPa, 290 °C) in low neutron dose (∼0.16 dpa, 185 °C) material. The TEM observations are utilized to design finite element (FE) meshes that include one or two “soft” channels (i.e. low critical resolved shear stress (CRSS)) of particular aspect ratio (length divided by thickness) embedded at the free surface of a “hard” matrix (i.e. high CRSS). The CRSS are adjusted using experimental data and physically based models from the literature. For doses leading to hardening saturation, the computed surface slips are as high as 100% for an applied stress close to the yield stress, when the observed channel aspect ratio is used. Surface slips are much higher than the grain boundary slips because of matrix constraint effect. The matrix CRSS and the channel aspect ratio are the most influential model parameters. Predictions based on an analytical formula are compared with surface slips computed by the FE method. Predicted slips, either in surface or bulk channels, agree reasonably well with either atomic force microscopy measures reported in the literature or measures based on our TEM observations. Finally, it is shown that the induced surface slip and grain boundary stress concentrations strongly enhance the kinetics of the damage mechanisms possibly involved in IASCC.     

The dependence of the high temperature mechanical properties of austenitic stainless steels on implanted helium

High temperature helium embrittlement effects on the creep properties of AISI 316 SS (solution annealed + aged) and DIN 1.4970 SS (solution annealed + cold worked + aged) have been investigated. The generation of helium due to (n. α) nuclear reactions in a fusion reactor environment has been simulated by homogeneous helium implantation at a cyclotron. The creep rupture tests with various applied tensile stresses have been carried out at 1023 K. (316 SS) and 1079 K (1.4970 SS). respectively, with four differently treatly sets of samples: (1) unimplanted controls; (2) after room temperature implantation of 100 appm He; (3) after implantation of 100 appm He at test temperature; (4) creep tested at high temperature during implantation (“in-beam”) with implantation rates of 10–100 appm He/b. In contrast to the ductile behaviour with transgranular failure of the unimplanted controls, all He-implanted samples showed brittle, intergranular early failure. The embrittlement effect was enhanced for the “in-beam” tested samples. The difference between the different treated sets of samples can be related to different bubble microstructures investigated by TEM. In addition, a comparison to reactor data for the DIN 1.4970 SS is presented.     

The effect of dose rate on the response of austenitic stainless steels to neutron radiation

Depending on reactor design and component location, austenitic stainless steels may experience significantly different irradiation dose rates in the same reactor. Understanding the effect of dose rate on radiation performance is important to predicting component lifetime. This study examined the effect of dose rate on swelling, grain boundary segregation, and tensile properties in austenitic stainless steels through the examination of components retrieved from the Experimental Breeder Reactor-II (EBR-II) following its shutdown. Annealed 304 stainless steel, stress-relieved 304 stainless steel, 12% cold-worked 316 stainless steel, and 20% cold-worked 316 stainless steel were irradiated over a dose range of 1–56 dpa at temperatures from 371 to 440 °C and dose rates from 0.5 to 5.8 × 10⁻⁷ dpa/s. Density and tensile properties were measured for 304 and 316 stainless steel. Changes in grain boundary composition were examined for 304 stainless steel. Swelling appears to increase at lower dose rates in both 304 and 316 stainless steel, although the effect was not always statistically significant. Grain boundary segregation also appears to increase at lower dose rate in 304 stainless steel. For the range of dose rates examined, no measurable dose rate effect on tensile properties was noted for any of the steels.     

Atomic scale investigation of radiation-induced segregation in austenitic stainless steels

Radiation-induced precipitation and segregation in a cold-worked 316 austenitic stainless steel irradiated with 10 MeV Fe⁵⁺ ions were characterized by atom probe tomography. Ni and Si enrichment and Cr depletion were observed in roughly spherical and torus-shaped clusters, believed to be due to solute enrichment and depletion at dislocation loops. Solute segregation was also observed at network dislocations. These observations are consistent with the phenomenon of radiation-induced segregation. Radiation-induced segregation at grain boundaries was also studied at the near atomic scale. Comparison of these observations with results from the literature shows a difference in the magnitude of the peak concentration of segregated solutes.     

Welding-induced microstructure in austenitic stainless steels before and after neutron irradiation

The effects of neutron irradiation on the microstructure of welded joints made of austenitic stainless steels have been investigated. The materials were welded AISI 304 and AISI 347, so-called test weld materials, and irradiated with neutrons at 300 °C to 0.3 and 1.0 dpa. In addition, an AISI 304 type from a decommissioned pressurised water reactor, so-called in-service material, which had accumulated a maximum dose of 0.35 dpa at about 300 °C, was investigated. The microstructure of heat-affected zones and base materials was analysed before and after irradiation, using transmission electron microscopy. Neutron diffraction was performed for internal stress measurements. It was found that the heat-affected zone contains, relative to the base material, a higher dislocation density, which relates well to a higher residual stress level and, after irradiation, a higher irradiation-induced defect density. In both materials, the irradiation-induced defects are of the same type, consisting in black dots and Frank dislocation loops. Careful analysis of the irradiation-induced defect contrast was performed and it is explained why no stacking fault tetrahedra could be identified.     

Irradiation damage in 304 and 316 stainless steels: experimental investigation and modeling. Part I: Evolution of the microstructure

Irradiation damage in three austenitic stainless steels, SA 304L, CW 316 and CW Ti-modified 316, is investigated both experimentally and theoretically. The density and size of Frank loops after irradiation at 320 and 375 °C in experimental EBR II, BOR-60 and OSIRIS reactors for doses up to 40 dpa are characterized by TEM. The evolution of the initial dislocation network under irradiation is evaluated. A cluster dynamics model is proposed to account quantitatively for the experimental findings.     

Effects of oversized solutes on radiation-induced segregation in austenitic stainless steels

Zirconium or hafnium additions to austenitic stainless steels caused a reduction in grain boundary Cr depletion after proton irradiations for up to 3 dpa at 400 °C and 1 dpa at 500 °C. The predictions of a radiation-induced segregation (RIS) model were also consistent with experiments in showing greater effectiveness of Zr relative to Hf due to a larger binding energy. However, the experiments showed that the effectiveness of the solute additions disappeared above 3 dpa at 400 °C and above 1 dpa at 500 °C. The loss of solute effectiveness with increasing dose is attributed to a reduction in the amount of oversized solute from the matrix due to growth of carbide precipitates. Atom probe tomography measurements indicated a reduction in amount of oversized solute in solution as a function of irradiation dose. The observations were supported by diffusion analysis suggesting that significant solute diffusion by the vacancy flux to precipitate surfaces occurs on the time scales of proton irradiations. With a decrease in available solute in solution, improved agreement between the predictions of the RIS model and measurements were consistent with the solute-vacancy trapping process, as the mechanism for enhanced recombination and suppression of RIS.     

Welding-induced mechanical properties in austenitic stainless steels before and after neutron irradiation

The effects of neutron irradiation on the mechanical properties of welded joints made of austenitic stainless steels have been investigated. The materials are welded AISI 304 and AISI 347, so-called test weld materials, irradiated with neutrons at 573 K to doses of 0.3 and 1.0 dpa. In addition, an AISI 304 from a decommissioned pressurised water reactor, so-called in-service material, which had accumulated a maximum dose of 0.35 dpa at about 573 K, was investigated. The mechanical properties of heat-affected zones and base materials were analysed before and after irradiation. Tensile parameters were determined at room temperature and at 573 K, for all materials and irradiation conditions. In the test weld materials it is found that radiation hardening is lower and loss of ductility is higher in the heat-affected zone than in the base material. In the in-service material radiation hardening is about the same in heat-affected zone and base material. After irradiation, deformation takes place by stacking faults and twins, at both room temperature and high temperature, contrary to unirradiated materials, where deformation takes place by twinning at room temperature and by dislocation cells at high temperature. No defect free channels are observed.     

Effects of nitrogen on low-cycle fatigue properties of type 304L austenitic stainless steels tested with and without tensile strain hold

A quantitative analysis of the effects of nitrogen on high temperature low-cycle fatigue without and with tensile strain hold at 600 °C has been conducted for type 304L stainless steels. For better understanding of the role of nitrogen on grain boundary precipitation, the grain boundary segregation of nitrogen was analyzed by Auger electron spectroscopy. The nitrogen addition is found to give relatively better resistance to creep-fatigue than continuous low-cycle fatigue. This in turn improves the fatigue life. This is due to the retardation of the precipitation of carbides at the grain boundary and reduction in the density of grain boundary cavitation sites which are the main factor of grain boundary damage under creep-fatigue test.     

Stress corrosion cracking susceptibility of austenitic stainless steels in supercritical water conditions

The presented paper summarizes the results of general corrosion and stress corrosion cracking (SCC) susceptibility tests in supercritical water (SCW), studied for austenitic stainless steel 316L, with the aim to identify maximum SCW temperature usability and specific failure mechanisms prevailing during slow strain-rate tensile (SSRT) tests in ultra-pure demineralized SCW solution with controlled oxygen content. The general corrosion tests clearly revealed the applicability of austenitic stainless steel in SCW to be limited to 550 °C as maximum temperature as oxidation rates of austenitic stainless steels 316L increase dramatically above 550 °C. The SSRT tests were performed using a step-motor controlled loading device in an autoclave at 550 °C SCW. Besides the strain rate (resp. crosshead speed), the oxygen content was varied in the series of tests. The obtained results showed that even at the lowest strain rate, a serious increase of SCC susceptibility, as typically characterized by IGSCC crack growth, was not observed. The fractography confirmed that failure was due to a combination of transgranular SCC and transgranular ductile fracture. Based on fractographic findings a phenomenological map describing the SCC regime of SSRT test parameters could be proposed for AISI 316L.     

Importance of strain rate in elevated-temperature low-cycle fatigue of austenitic stainless steels

Extrapolation of elevated-temperature, tensile-hold fatigue life of types 304 and 316 stainless steel is obtained by the use of four existing life predictive methods. The results show that, although the calculated lives for the different methods are similar for short hold-time tests, they can vary greatly from one method to another when extrapolated to long hold-time situations. Methods that do not take into account the effects of strain rate provide optimistic values as opposed to the more pessimistic values projected by the methods that account for strain-rate effects.     

Irradiation creep of SA 304L and CW 316 stainless steels: Mechanical behaviour and microstructural aspects. Part II: Numerical simulation and test of SIPA model

A cluster dynamic model has been adapted to test the Stress Induced Preferential Absorption of Defect (SIPA) on Frank loops hypothesis concerning irradiation creep, to reproduce quantitatively both microstructure evolution and its stress induced anisotropy and macroscopic creep rate. It is concluded that SIPA on Frank loops model can account for the observed defects structure, but is unable to reproduce quantitatively the creep rate.