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.     

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.     

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.     

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.     

Deformation and microstructure of neutron irradiated stainless steels with different stacking fault energy

The influence of the stacking fault energy (SFE) on the microstructure, mechanical property and deformation behaviour of stainless steels before and after irradiation was investigated. The mechanical properties, such as strength, ductility, strain hardening and irradiation induced hardening behaviours of 3 alloys with various SFEs are quite different. Such significant variations of mechanical properties must result from the different microstructures, deformation mechanisms and defects accumulation behaviours. Thus, the microstructures, deformation mechanisms and irradiation induced small defect clusters (including their types, natures, densities and size distributions) of 3 alloys are determined in detail by transmission electron microscopy. It indicated that before irradiation, alloy with low SFE has more localised deformation behaviour than alloy with high SFE. After irradiation, in the samples with low SFE, the irradiation induced stacking fault tetrahedral was observed, while in the ones with high SFE, the dominant defects are Frank loops. All the results shown that, SFE has a strong effect on both the deformation mechanism and irradiation induced defect accumulation behaviour of stainless steels.     

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.     

The microstructure of neutron irradiated 20Cr/25Ni/TiN austenitic stainless steel

A nitrided, titanium-stabilised, 20Cr/25Ni austenitic stainless steel was examined in the electron microscope, after 1000 h irradiation at 783 K to doses of 5.0× 10²⁴ n/m² (thermal) and 2.5× 10²⁴ n/m² (fast), in the PLUTO reactor. Microstructures were compared with those of as-received and thermal control samples. The austenitic matrix and M₂₃C₆ particles were free of irradiation-induced damage, while the TiN particles contained loops 2 nm in diameter which coarsened into a network on annealing at 1083 K. Annealing also resulted in a low density of transmutation-induced helium bubbles, ~ 4 nm in diameter, located in precipitate-free regions of grain boundaries. We conclude that 20/25/TiN is relatively unaffected by irradiation at these dose levels and that helium bubble embrittlement is unlikely under normal stresses.     

The effect of fast neutron irradiation upon the fatigue-crack propagation behavior of two austenitic stainless steels

The effect of fast neutron irradiation ($\text{454° < T}{}_{\mathrm{irr}}\text{< 477° C}$) to a fluence of $\text{9 × 10}{}^{21}\text{n/cm}{}^2$ ($\text{E > 0.1 MeV}$) on the fatigue-crack growth behavior was investigated for annealed Type 304 and 20% coldworked Type 316 stainless steels using linear-elastic fracture mechanics techniques. Irradiation to this fluence had little or no effect upon the crack growth behavior of annealed Type 304 at a test temperature of 427° C, nor upon the behavior of 20% cold-worked Type 316 at test temperatures of 427° C and 538° C. Irradiation to this fluence did tend to decrease crack growth rates slightly, relative to unirradiated material, in annealed Type 304 at a test temperature of 538° C.     

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.     

The effect of prior cold-work on the deformation behaviour of neutron irradiated AISI 304 austenitic stainless steel

Cold-work is intentionally employed to increase the yield strength of austenitic stainless steels and also occurs during fabrication processes, but it has also been associated with greater incidence of stress corrosion cracking. This study examined the effect of up to 3.85 dpa neutron irradiation on the deformation behaviour and microstructures of 30% cold-worked AISI 304 material tensile tested at 300 °C. While the deformation behaviour of 0.07 dpa material was similar to non-irradiated material tested at the same temperature, its stress-strain curve was shifted upwards by about 200 MPa. Materials irradiated to over 2 dpa hardened some 400-500 MPa, but showed limited strain hardening capacity, exhibiting precipitous softening with further straining beyond the yield point. The observed behaviour is most likely a consequence of planar deformation products serving as strengtheners to the unirradiated bulk on the one hand, while promoting strain localization on the other, behaviour exacerbated by the subsequent neutron irradiation.     

Fracture mechanics evaluations of neutron irradiated Type 321 austenitic steel

The effects of fast reactor irradiation at temperatures of ~ 230° C and ~ 400° C on the fracture toughness and associated strength changes, induced in solution treated Type 321 stainless steel have been characterised using instrumented impact test procedures. The studies cover irradiation exposures in the range 16 to 43 displacements per atom (dpa) and test temperatures of 23–500° C.Irradiation results in significant but not catastrophic reductions in fracture toughness, together with radiation hardening effects. Both the dose and test temperature dependence of the toughness changes are sensitive to irradiation temperature. Thus, whilst maximum toughness loss occurs at or below 16 dpa for both irradiation temperatures, the 400° C-irradiation condition is associated with subsequent saturation of the toughness change, whereas for 230° C-irradiation measurable but low on-going toughness degradation occurs up to 43 dpa. The fracture toughness characteristics correlate with fractographic observations which demonstrate retention of a predominantly ductile fracture mode after irradiation, but with dramatic refinement in the scale of microvoid coalescence associated with TiC precipitates. It is suggested that the fracture mechanism after irradiation is controlled primarily by the irradiation-induced precipitate distribution, and furthermore, that the maintenance of ductile fracture, and hence good toughness, up to high irradiation damage levels is a consequence of inhibition of incipient channel fracture processes by the TiC particles.The application of general yield fracture mechanics to calculate critical defect sizes for unstable fracture in fast reactor wrappers is illustrated. These assessments demonstrate the importance of considering net section yield as an alternative failure criterion in thin section components. Finally, the use of empirical equations as a design philosophy to predict irradiation-induced toughness changes is briefly considered.     

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.     

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.     

Dynamic uniaxial and biaxial stress-strain relationships for austenitic stainless steels

The question of reliability technology using quantified techniques is considered for systems and structures. Systems reliability analysis has progressed to a viable and proven methodology whereas this has yet to be fully achieved for large scale structures.Structural loading variants over the life-time of the plant are considered to be more difficult to analyse than for systems, even though a relatively crude model may be a necessary starting point. Various reliability characteristics and environmental conditions are considered which enter this problem.The rare event situation is briefly mentioned together with aspects of proof testing and normal and upset loading conditions.     

Influence of silicon on swelling and microstructure in Russian austenitic stainless steel EI-847 irradiated to high neutron doses

Void swelling and microstructural development of niobium-stabilized EI-847 austenitic stainless steel with a range of silicon levels were investigated by destructive examination of fuel pin cladding irradiated in three fast reactors located in either Russia or Kazakhstan. The tendency of void swelling to be progressively reduced by increasing silicon concentration appears to be a very general phenomenon in this steel, whether observed in simple, single-variable experiments on well-defined materials or when observed in multivariable, time-dependent irradiations conducted on commercially produced steels over a wide range of irradiation temperatures, neutron spectra and dpa rates. The role of silicon on microstructural development is expressed both in the solid solution via its influence on dislocation and void microstructure and via its influence on formation of radiation-induced phases that in turn alter the matrix composition. Surprisingly, increases in silicon level in this study do not accelerate the formation of silicon-rich G-phase, but act to increase the formation of Nb (C,N) precipitates. Such precipitates are known to be associated with delayed void swelling.     

Dislocation loop evolution under ion irradiation in austenitic stainless steels

A solution annealed 304 and a cold worked 316 austenitic stainless steels were irradiated from 0.36 to 5 dpa at 350 °C using 160 keV Fe ions. Irradiated microstructures were characterized by transmission electron microscopy (TEM). Observations after irradiation revealed the presence of a high number density of Frank loops. Size and number density of Frank loops have been measured. Results are in good agreement with those observed in the literature and show that ion irradiation is able to simulate dislocation loop microstructure obtained after neutron irradiation.Experimental results and data from literature were compared with predictions from the cluster dynamic model, MFVIC (Mean Field Vacancy and Interstitial Clustering). It is able to reproduce dislocation loop population for neutron irradiation. Effects of dose rate and temperature on the loop number density are simulated by the model. Calculations for ion irradiations show that simulation results are consistent with experimental observations. However, results also show the model limitations due to the lack of accurate parameters.     

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.     

Understanding silicon-rich phase precipitation under irradiation in austenitic stainless steels

Atom probe samples have been Fe⁺ ion irradiated at different doses (from 0.5 to 10 dpa) and different temperatures (between 300 and 400 °C) in order to understand the mechanism of formation, under irradiation, of Si-rich phases in austenitic stainless steels. Atom probe results show the presence of Si-enriched clusters which can also be enriched in Ni and depleted in Cr. Number densities of solute clusters can be linked to number densities of dislocation loops already observed by transmission electron microscopy in a previous work. This suggests that solute clusters are formed by heterogeneous precipitation on dislocation loops. Furthermore, the evolution of the composition of solute clusters as a function of the irradiation temperature is consistent with a radiation-induced mechanism. Results are also compared with previous results obtained after neutron irradiation at lower dose rate (in term of dpa s⁻¹). The comparison is, here again, consistent with the radiation-induced mechanism. Thus, Si-rich clusters may be formed by radiation-induced segregation to dislocation loops. Results also show that Si is probably dragged to sinks via the interstitial mechanism.     

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.