Weldability of neutron irradiated austenitic stainless steels

Degradation of weldability in neutron irradiated austenitic stainless steel is an important issue to be addressed in the planning of proactive maintenance of light water reactor core internals. In this work, samples selected from reactor internal components which had been irradiated to fluence from 8.5 × 10²² to 1.4 × 10²⁶ n/m² (E > 1 MeV) corresponding to helium content from 0.11 to 103 appm, respectively, were subjected to tungsten inert gas arc (TIG) welding with heat input ranged 0.6–16 kJ/cm. The weld defects were characterized by penetrant test and cross-sectional metallography. The integrity of the weld was better when there were less helium and at lower heat input. Tensile properties of weld joint containing 0.6 appm of helium fulfilled the requirement for unirradiated base metal. Repeated thermal cycles were found to be very hazardous. The results showed the combination of material helium content and weld heat input where materials can be welded with little concern to invite cracking. Also, the importance of using properly selected welding procedures to minimize thermal cycling was recognized.     

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.     

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.     

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.     

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.     

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.     

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.     

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.     

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.     

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 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.     

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.     

Corrosion mechanisms of austenitic stainless steels in nitric media used in reprocessing plants

Austenitic stainless steels type 304L, 316L and 310Nb are largely used as structural materials for equipments handling nitric acid media in reprocessing plants. In almost all nitric media, these materials, protected by a chromium(III) oxide rich layer, remain in their passive state. However, in some particular nitric media, their corrosion potential may be shifted towards their transpassive domain. In this domain, they can suffer intergranular corrosion, even though they are not sensitized owing to their very low carbon content. The corrosion potential of the steel depends greatly on the cathodic reaction involved in the oxido-reduction process between the elements Fe, Cr, Ni of the steel and the oxidizing species of the medium. Three cases of an increase in the corrosion potential can be found in reprocessing media: pure nitric acid-water solutions, in which the cathodic reaction is the reduction reaction of HNO₃; nitric acid media containing oxidizing species, in which the cathodic reaction is the reaction of reduction of the oxidizing species into the reduced one; nitric media containing metallic elements electrochemically more noble than the steels, causing galvanic coupling. In each case, the mechanism and the relevant situations we experimentally studied are described.     

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.     

Time-temperature-sensitization diagrams and critical cooling rates of different nitrogen containing austenitic stainless steels

Nitrogen-alloyed 316L stainless steel is being used as structural material for high temperature fast breeder reactor components with a design life of 40 years. With a view to increase the design life to 60 years and beyond, high nitrogen stainless steels are being considered for certain critical components which may be used at high temperatures. Since carbon and nitrogen have major influence on the sensitization kinetics, investigations were carried out to establish the sensitization behaviour of four heats of 316L SS containing (i) 0.07%N and 0.035%C, (ii) 0.120%N and 0.030%C, (iii) 0.150%N and 0.025%C and (iv) 0.22%N and 0.035%C. These stainless steels were subjected to heat treatments in the temperature range of 823-1023 K for various durations ranging from 1 h to 500 h. Using ASTM standard A262 Practice A and E tests, time-temperature-sensitization diagrams were constructed and from these diagrams, critical cooling rate above which there is no risk of sensitization was calculated. The data established in this work can be used to select optimum heat treatment parameters during heat treatments of fabricated components for fast reactors.     

Formulation of a new intergranular creep damage model for austenitic stainless steels

The aim of this study is to propose a physically based intergranular creep damage model for numerical simulations on extrapolated situations. A continuum damage formulation is proposed to evaluate nucleation, growth and coalescence of intergranular creep cavities. Nucleation is based on an empirical law where void fraction growth rate is proportional to the creep strain rate. Void growth rate includes the contribution of: viscoplastic strain rate of surrounding grains (Gurson), and vacancy diffusion along grain boundaries (Hull and Rimmer). Void coalescence is based on a mechanical fracture criterion, where the competition between damage softening and viscoplastic hardening is considered. The identification procedure needs only the results of uniaxial creep tensile tests with a range of time to rupture that enables a sufficient diffusion contribution. The constraint effect is taken into account in the formulation of the model and does not need a specific identification. To illustrate the capacity of the proposed model, applications are presented for an austenitic stainless steel tested at 600 °C. It appears that the constraint effect assessment is in good agreement with experimental results, when we compare time to rupture and intergranular damage localisation on notched specimens, or crack initiation time and crack growth rate on fatigue pre-cracked specimens.     

Solute segregation along non-migrated and migrated grain boundaries during electron irradiation in austenitic stainless steels

Solute distribution and precipitation in the vicinity of the grain boundary in Type 316 steels were studied during electron irradiation up to about 50 dpa at temperatures from room temperature to 873 K. Undersized solute atoms, such as nickel, silicon and phosphorus, segregate toward the grain boundary, and oversized solutes, chromium and molybdenum, segregate away from the grain boundary during irradiation in the temperature range between 623 and 873 K. Enrichment of silicon and phosphorus along the grain boundary occurs after the irradiation at room temperature. The segregation of solute atoms increases with irradiation temperature except for silicon and phosphorus; the concentration of silicon and phosphorus along the grain boundary exhibits a maximum at 773 K. Remarkable depletion of chromium with enrichment of nickel, silicon and phosphorus occurs in the area swept by the migrating grain boundary. Massive M₂₃C₆ type carbide precipitates in front of the migrating grain boundary during irradiation in the temperature range from 723 to 873 K in the steels.     

Interpretation of the influences of irradiation upon fatigue crack propagation in austenitic stainless steels

An interpretation of the influences of neutron irradiation upon fatigue crack propagation in austenitic stainless steels is given. The approach has been to extend a previously developed rationalisation of the effects of various test and materials variables upon fatigue crack propagation in unirradiated stainless steels to include irradiated stainless steels.Irradiation has diverse influences upon the rate of fatigue crack propagation depending on the exact irradiation and test conditions. It has been shown that, by considering the underlying mechanisms of failure, some confidence is established in trends in data in a subject where information is very scarce and difficult to obtain.