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.     

Proton irradiation emulation of PWR neutron damage microstructures in solution annealed 304 and cold-worked 316 stainless steels

Solution annealed (SA) 304 and cold-worked (CW) 316 austenitic stainless steels were pre-implanted with helium and were irradiated with protons in order to study the potential effects of helium, irradiation dose, and irradiation temperature on microstructural evolution, especially void swelling, with relevance to the behavior of austenitic core internals in pressurized water reactors (PWRs). These steels were irradiated with 1 MeV protons to doses between 1 and 10 dpa at 300 °C both with or without 15 appm helium pre-implanted at ∼100 °C. They were also irradiated at 340 °C, but only after 15 appm helium pre-implantation. Small heterogeneously distributed voids were observed in both alloys irradiated at 300 °C, but only after helium pre-implantation. The pre-implanted steels irradiated at 340 °C exhibited homogenous void formation, suggesting effects of both helium and irradiation temperature on void nucleation. Voids developed sooner in the SA304 alloy than CW316 alloy at 300 and 340 °C, consistent with the behavior observed at higher temperatures (>370 °C) for similar steels irradiated in the EBR-II fast reactor. The development of the Frank loop microstructure was similar in both alloys, and was only marginally affected by pre-implanted helium. Loop densities were insensitive to dose and irradiation temperature, and were decreased by helium; loop sizes increased with dose up to about 5.5 dpa and were not affected by the pre-implanted helium. Comparison with microstructures produced by neutron irradiation suggests that this method of helium pre-implantation and proton irradiation emulates neutron irradiation under PWR conditions.     

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.     

Tensile properties of EC316LN irradiated in SINQ to 20 dpa

The wide application of 316-type austenitic stainless steels in existing spallation targets requires a comprehensive understanding of their behavior in spallation irradiation environments. In the present study, EC316LN specimens were irradiated in SINQ targets to doses between 3 and 17.3 dpa at temperatures between about 80 °C and 390 °C. Tensile tests were conducted at room and irradiation temperatures. The results demonstrate that the irradiation induced significant hardening and embrittlement in the specimens. The irradiation hardening and embrittlement effects show a trend of saturation at doses above about 10 dpa. Although the ductility was greatly reduced, all specimens broke with strong necking, which indicates a ductile fracture mode.     

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.     

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.     

Irradiation of structural materials in contact with lead bismuth eutectic in the high flux reactor

In the framework of the materials domain DEMETRA in the European Transmutation research and development project EUROTRANS, irradiation experiment IBIS has been performed in the High Flux Reactor in Petten. The objective was to investigate the synergystic effects of irradiation and lead bismuth eutectic exposure on the mechanical properties of structural materials and welds. In this experiment ferritic martensitic 9 Cr steel, austenitic 316L stainless steel and their welds have been irradiated for 250 Full Power Days up to a dose level of 2 dpa. Irradiation temperatures have been kept constant at 300 °C and 500 °C.During the post-irradiation test phase, tensile tests performed on the specimens irradiated at 300 °C have shown that the irradiation hardening of ferritic martensitic 9 Cr steel at 1.3 dpa is 254 MPa, which is in line with the irradiation hardening obtained for ferritic martensitic Eurofer97 steel investigated in the fusion program. This result indicates that no LBE interaction at this irradiation temperature is present. A visual inspection is performed on the specimens irradiated in contact with LBE at 500 °C and have shown blackening on the surface of the specimens and remains of LBE that makes a special cleaning procedure necessary before post-irradiation mechanical testing.     

Irradiation-Induced Deformation in Cold-Worked SUS 304 Stainless Steel

SUS 304 stainless steel has been used in the light-water reactors constructed in earlier days, in which irradiation-assisted stress corrosion cracking has drawn increasing attention and tensile residual stress is believed to be one of the major causes. It is, therefore, essential to assess its stress relaxation behavior under irradiation, which can be evaluated from the irradiation creep data, and the effect of cold work on it. Creep experiments under 17 MeV proton irradiation (2x10^?7 dpa/s) at 288°C were conducted for SUS 304 with 5% and 25% cold work (CW). Irradiation creep rate of 5%CW was only slightly larger than that of 25%CW. Stress dependence was almost quadratic in both specimens, in contrast with the linear dependence in cold-worked SUS 316L reported earlier. Stress relaxation under irradiation was found to reflect this quadratic dependence. Martensite is induced by cold-working in SUS 304, not in SUS 316L, and marked difference in its amount was found between 5%CW and 25%CW, despite the small difference in irradiation creep behavior. Thus, the observed quadratic dependence appears to result not directly from the induced martensite itself but from a climb-enabled glide of the tangled dislocations densely formed in the vicinity of martensite phase boundaries.     

Radiation-induced stress relaxation in high temperature water of type 316L stainless steel evaluated by neutron diffraction

Weld beads on plate specimens made of type 316L stainless steel were neutron-irradiated up to about 2.5 × 10²⁵ n/m² (E > 1 MeV) at 561 K in the Japan Material Testing Reactor (JMTR). Residual stresses of the specimens were measured by the neutron diffraction method, and the radiation-induced stress relaxation was evaluated. The values of σ_x residual stress (transverse to the weld bead) and σ_y residual stress (longitudinal to the weld bead) decreased with increasing neutron dose. The tendency of the stress relaxation was almost the same as previously published data, which were obtained for type 304 stainless steel. From this result, it was considered that there was no steel type dependence on radiation-induced stress relaxation. The neutron irradiation dose dependence of the stress relaxation was examined using an equation derived from the irradiation creep equation. The coefficient of the stress relaxation equation was obtained, and the value was 1.4 (×10⁻⁶/MPa/dpa). This value was smaller than that of nickel alloy.     

Analysis of tensile deformation and failure in austenitic stainless steels: Part II - Irradiation dose dependence

Irradiation effects on the stable and unstable deformation and fracture behavior of austenitic stainless steels (SSs) have been studied in detail based on the equivalent true stress versus true strain curves. An iterative finite element simulation technique was used to obtain the equivalent true stress-true strain data from experimental tensile curves. The simulation result showed that the austenitic stainless steels retained high strain hardening rate during unstable deformation even after significant irradiation. The strain hardening rate was independent of irradiation dose up to the initiation of a localized necking. Similarly, the equivalent fracture stress was nearly independent of dose before the damage (embrittlement) mechanism changed. The fracture strain and tensile fracture energy decreased with dose mostly in the low dose range <∼2 dpa and reached nearly saturation values at higher doses. It was also found that the fracture properties for EC316LN SS were less sensitive to irradiation than those for 316 SS, although their uniform tensile properties showed almost the same dose dependencies. It was confirmed that the dose dependence of tensile fracture properties evaluated by the linear approximation model for nominal stress was accurate enough for practical use without elaborate calculations.     

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.     

Temperature effects on the mechanical properties of candidate SNS target container materials after proton and neutron irradiation

This report presents the tensile properties of EC316LN austenitic stainless steel and 9Cr-2WVTa ferritic/martensitic steel after 800 MeV proton and spallation neutron irradiation to doses in the range 0.54-2.53 dpa at 30-100 °C. Tensile testing was performed at room temperature (20 °C) and 164 °C. The EC316LN stainless steel maintained notable strain-hardening capability after irradiation, while the 9Cr-2WVTa ferritic/martensitic steel posted negative hardening in the engineering stress-strain curves. In the EC316LN stainless steel, increasing the test temperature from 20 to 164 °C decreased the strength by 13-18% and the ductility by 8-36%. The effect of test temperature for the 9Cr-2WVTa ferritic/martensitic steel was less significant than for the EC316LN stainless steel. In addition, strain-hardening behaviors were analyzed for EC316LN and 316L stainless steels. The strain-hardening rate of the 316 stainless steels was largely dependent on test temperature. A calculation using reduction of area measurements and stress-strain data predicted positive strain hardening during plastic instability.     

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.     

Isolating the effect of radiation-induced segregation in irradiation-assisted stress corrosion cracking of austenitic stainless steels

Post-irradiation annealing was used to help identify the role of radiation-induced segregation (RIS) in irradiation-assisted stress corrosion cracking (IASCC) by preferentially removing dislocation loop damage from proton-irradiated austenitic stainless steels while leaving the RIS of major and minor alloying elements largely unchanged. The goal of this study is to better understand the underlying mechanisms of IASCC. Simulations of post-irradiation annealing of RIS and dislocation loop microstructure predicted that dislocation loops would be removed preferentially over RIS due to both thermodynamic and kinetic considerations. To verify the simulation predictions, a series of post-irradiation annealing experiments were performed. Both a high purity 304L (HP-304L) and a commercial purity 304 (CP-304) stainless steel alloy were irradiated with 3.2 MeV protons at 360 °C to doses of 1.0 and 2.5 dpa. Following irradiation, post-irradiation anneals were performed at temperatures ranging from 400 to 650 °C for times between 45 and 90 min. Grain boundary composition was measured using scanning transmission electron microscopy with energy-dispersive spectrometry in both as-irradiated and annealed samples. The dislocation loop population and radiation-induced hardness were also measured in as-irradiated and annealed specimens. At all annealing temperatures above 500 °C, the hardness and dislocation densities decreased with increasing annealing time or temperature much faster than RIS. Annealing at 600 °C for 90 min removed virtually all dislocation loops while leaving RIS virtually unchanged. Cracking susceptibility in the CP-304 alloy was mitigated rapidly during post-irradiation annealing, faster than RIS, dislocation loop density or hardening. That the cracking susceptibility changed while the grain boundary chromium composition remained essentially unchanged indicates that Cr depletion is not the primary determinator for IASCC susceptibility. For the same reason, the visible dislocation microstructure and radiation-induced hardening are also not sufficient to cause IASCC alone.     

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.     

Emulation of neutron irradiation effects with protons: validation of principle

This paper presents the results of the irradiation, characterization and irradiation assisted stress corrosion cracking (IASCC) behavior of proton- and neutron-irradiated samples of 304SS and 316SS from the same heats. The objective of the study was to determine whether proton irradiation does indeed emulate the full range of effects of in-reactor neutron irradiation: radiation-induced segregation (RIS), irradiated microstructure, radiation hardening and IASCC susceptibility. The work focused on commercial heats of 304 stainless steel (heat B) and 316 stainless steel (heat P). Irradiation with protons was conducted at 360 °C to doses between 0.3 and 5.0 dpa to approximate those by neutron irradiation at 275 °C over the same dose range. Characterization consisted of grain boundary microchemistry, dislocation loop microstructure, hardness as well as stress corrosion cracking (SCC) susceptibility of both un-irradiated and irradiated samples in oxygenated and de-oxygenated water environments at 288 °C. Overall, microchemistry, microstructure, hardening and SCC behavior of proton- and neutron-irradiated samples were in excellent agreement. RIS analysis showed that in both heats and for both irradiating particles, the pre-existing grain boundary Cr enrichment transformed into a ‘W' shaped profile at 1.0 dpa and then into a ‘V' shaped profile between 3.0 and 5.0 dpa. Grain boundary segregation of Cr, Ni, Si, and Mo all followed the same trends and agreed well in magnitude. The microstructure of both proton- and neutron-irradiated samples was dominated by small, faulted dislocation loops. Loop size distributions were nearly identical in both heats over a range of doses. Saturated loop size following neutron irradiation was about 30% larger than that following proton irradiation. Loop density increased with dose through 5.0 dpa for both particle irradiations and was a factor of 3 greater in neutron-irradiated samples vs. proton-irradiated samples. Grain boundary denuded zones were only observed in neutron-irradiated samples. No cavities were observed for either irradiating particle. For both irradiating particles, hardening increased with dose for both heats, showing a more rapid increase and approach to saturation for heat B. In normal oxygenated water chemistry (NWC) at 288 °C, stress corrosion cracking in the 304 alloy was first observed at about 1.0 dpa and increased with dose. The 316 alloy was remarkably resistant to IASCC for both particle types. In hydrogen treated, de-oxygenated water (HWC), proton-irradiated samples of the 304 alloy exhibited IG cracking at 1.0 dpa compared to about 3.0 dpa for neutron-irradiated samples, although differences in specimen geometry, test condition and test duration can account for this difference. Cracking in heat P in HWC occurred at about 5.0 dpa for both irradiating particles. Thus, in all aspects of radiation effects, including grain boundary microchemistry, dislocation loop microstructure, radiation hardening and SCC behavior, proton-irradiation results were in good agreement with neutron-irradiation results, providing validation of the premise that the totality of neutron-irradiation effects can be emulated by proton irradiation of appropriate energy.     

Irradiation effect on mechanical properties in structural materials of fast breeder reactor plant

The effects of displacement per atom (dpa) level, helium content, and the ratio of helium content to dpa level on the tensile and creep properties have been investigated in the assumed irradiation damage range of FBR structural materials. The assumed irradiation damage range is up to about 1 dpa and about 30 appm for helium content. Austenitic stainless steel and high-chromium martensitic steel are considered as FBR structural materials. As a result, it is shown that the dpa level is a promising index for evaluating neutron irradiation damage.     

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.     

Effects of hold time and neutron irradiation on the low-cycle fatigue behaviour of type 316-CL and their consideration in a damage model

The creep fatigue behaviour of AISI type 316 L(N) plate material has been investigated in the temperature range of 450–750 °C by performing axial strain controlled tests with GRIM specimens. The creep and creep fatigue behaviour of austenitic stainless steel material is known to be prone to neutron irradiation-induced embrittlement. Therefore, the irradiation behaviour was studied by performing irradiation experiments in the High Flux Reactor (HFR) of Petten at 550 °C. A newly developed damage model for time-dependent damage was applied to describe the failure behaviour of AISI 316 L(N) in the cyclic tests performed.     

The effect of neutron irradiation on mechanical properties of YAG laser weldments using previously irradiated material

The objective of this study is to make clear the effect of neutron irradiation on mechanical properties of laser weldments using irradiated material. This estimation is necessary for the application to joining coolant piping of the ITER blanket. Irradiation testing was performed at Japan Material Testing Reactor (JMTR). On the irradiation condition for weldments using irradiated material, fast neutron fluence was 1.4 × 10²⁴ n/m², which corresponds to a displacement damage rate of 0.26 displacement per atom (dpa) and irradiation temperature 200 °C. The results of this study show that tensile properties of all weldments changed into that of base material by the effect of neutron irradiation. The results of hardness tests show that irradiation hardening at an irradiation damage dose of 0.3 dpa is almost same as that at irradiation damage 0.6 dpa. It is concluded that irradiated weldments using irradiated material were moved toward irradiated base material on tensile and hardness properties up to 0.6 dpa. On the other hand, tensile properties of base material were changed by the effect of neutron irradiation up to about 0.3 dpa, and with much less change from 0.3 dpa to 0.6 dpa. It is inferred that the effect of neutron irradiation of SS316LN-IG almost saturated up to 0.3 dpa.