Irradiated Austenitic Stainless Steels Research Articles

Radiation-induced segregation on dislocation loops in austenitic Fe-Cr-Ni alloys

Radiation-induced segregation of alloying elements to crystallographic defects is commonly observed in irradiated austenitic stainless steels. The interaction between solutes and radiation-induced defects changes the physical distribution of solutes and thus affects the formation and growth of defects. The change of the microstructure consequently affects the mechanical properties of the material. A qualitative and quantitative understanding of the interaction between solutes and defects is desirable to better predict the service lifetime of nuclear materials. We used atom probe tomography to measure the distribution of solutes at dislocation loops in 304L stainless steel, irradiated with 2 MeV protons up to 1.5 displacements per atom at 373 and 633 K. No segregation at dislocation loops was found in samples irradiated at 373 K, whereas Ni and Si enrichment and Cr depletion were detected at dislocation loops irradiated at 633 K. The experimentally observed perfect and faulted dislocation loops in vacancy and interstitial types were reproduced by molecular dynamics (MD). A hybrid MD/Monte Carlo method was used to predict the redistribution of alloying atoms at all possible types of dislocation loops in face-centered cubic Fe-Cr-Ni alloys at the same irradiated temperatures (373 and 633 K). The simulations show that, at both temperatures, Cr clusters were formed and distributed randomly, and Ni atoms enriched or depleted at interstitial or vacancy dislocation loops, respectively. The change of solute concentration reaches the highest at the edge of the loop. Ni profiles exhibit characteristic behavior in terms of the stress field of the loops: tension inside of vacancy loops showing depletion of Ni atoms compared with compression inside of interstitial loops showing enrichment of Ni atoms. In addition, the stress field is reduced after solute redistribution. The absence of alloying segregation observed in experiments at a lower temperature (373 K) is explained by a rate theory model: Low-temperature irradiation requires significantly longer irradiation time to see the same amount of segregation as at high temperatures because of the extremely low diffusion of vacancies at low temperatures.

Formulation of Irradiation Assisted Stress Corrosion Crack Growth Rates for Neutron-Irradiated Stainless Steels under Hydrogen Water Chemistry Conditions of Boiling Water Reactors

ABSTRACT This study developed a formula for the calculation of crack growth rates (CGRs) of irradiated austenitic stainless steels (SSs) under hydrogen water chemistry (HWC) conditions of boiling water reactors. First, the input parameters for the Hashimoto–Koshiishi model, which is a mechanistic crack growth model of stress corrosion cracking (SCC), were optimised for SSs irradiated up to 11 dpa. Electrochemical corrosion potential (ECP) of those SSs during crack growth tests nearly equalled −200 mVSHE. Second, a simplified formulation was developed to reproduce the model calculations of SSs. Calculated CGRs agreed well with experimental CGRs of various types of SSs, which were irradiated up to 13 dpa and tested at ECPs from −550 mVSHE to −200 mVSHE, when the experimental CGRs were less than the values for NWC conditions. On the other hand, there were some data in which experimental CGRs were much higher than calculated CGRs. By analysis of the experimental conditions of the data, the possible influencing factor for the discrepancy between the calculated and experimental CGRs was found to be the -validity. If the -validity criterion was not satisfied, the strain distribution used in the model calculation was not adequate for the actual strain distribution.

Probing the Damage Recovery Mechanism in Irradiated Stainless Steels Using In-Situ Microcantilever Bending Test

Single crystalline microcantilevers are fabricated from the base metal and heat-affected zone (HAZ) of a laser welded, neutron irradiated austenitic stainless steel, for scanning electron microscope (SEM) in-situ bending. In the HAZ, cantilevers exhibit higher yield point and lower crack tip blunting displacement than in the base metal and unirradiated archive specimen. These results suggest that radiation-induced defects harden the base metal, whereas the HAZ exhibits annealing of defects leading to mechanical softening. Dislocation nucleation ahead of the crack tip is responsible for ductile blunting behavior and provides a pathway to mitigating helium-induced cracking during weld repairs of irradiated materials.

A micromechanical analysis of intergranular stress corrosion cracking of an irradiated austenitic stainless steel

Irradiation Assisted Stress Corrosion Cracking (IASCC) is a material degradation phenomenon affecting austenitic stainless steels used in nuclear Pressurized Water Reactors (PWR), leading to the initiation and propagation of intergranular cracks. Such phenomenon belongs to the broader class of InterGranular Stress Corrosion Cracking (IGSCC). A micromechanical analysis of IGSCC of an irradiated austenitic stainless steel is performed in this study to assess local cracking conditions. A 304L proton irradiated sample tested in PWR environment and showing intergranular cracking is investigated. Serial sectioning, Electron BackScatter Diffraction (EBSD) and a two-step misalignment procedure are performed to reconstruct the 3D microstructure over an extended volume, to assess statistically cracking criteria. A methodology is also developed to compute Grain Boundary (GB) normal orientations based on the EBSD measurements. The statistical analysis shows that cracking occurs preferentially for GB normals aligned with the mechanical loading axis, but also for low values of the Luster-Morris slip transmission parameter. Micromechanical simulations based on the reconstructed 3D microstructure, FFT-based solver and crystal plasticity constitutive equations modified to account for slip transmission at grain boundaries are finally performed. These simulations rationalize the correlation obtained experimentally into a single stress-based criterion. The actual strengths and weaknesses of such micromechanical approach are discussed.

Irradiation damages of structural materials under different irradiation environments

For the advancement and development of nuclear systems used in heavy irradiation environments such as fusion DEMO reactors, fission reactors, fast reactors, and accelerator driven target systems, it is necessary to fully understand the changes of mechanical properties and the other properties of the materials induced by irradiation and to clarify the synergistic effect of displacement damage and helium generation. In this study the mechanical property changes and microstructural development induce by displacement damage and helium production have been mainly examined in austenitic stainless steels, 316FR and type304, and ferritic/martensitic steel, HCM12A, irradiated at around 550oC in JRR-3M reactor and/or JOYO fast reactor. At 550°C, 316FR steel was superior to 304 steel in terms of the amount of ductility and strength with respect to irradiation resistance. It is noteworthy that at 550 °C as well as room temperature, the higher fracture strength of the 316FR steels is a remarkable result. It is found that helium atoms strongly influenced on creep lifetime of the irradiated austenitic stainless steel, 316FR. It was found that the ratio of creep rupture time is slightly lower than the lower limit of previous study's Miyaji and co-workers in the region from 0.01 appm to 1 appm. It is also found that the lower limit of reduction ratio of creep rupture time (irradiation specimen to unirradiation one) does not decrease linearly with the helium production above 10 appm up to about 33 appm. dpa enhanced the reduction of creep lifetime. Recent R&D of high-energy accelerator driven target systems used under heavy irradiation environment is also introduced and discussed for high radiation resistance materials such as Ti alloys with very high number density of nano size precipitate, which have been studying under RaDIATE collaboration, from points of view of irradiation damage and materials development.

Technical Basis Part 2 for Code Case N-889: Reference Stress Corrosion Crack Growth Rate Curves for Irradiated Austenitic Stainless Steels in Light Water Reactor Environments

Abstract A multiyear international data collection, data review, modeling, and implementation project was recently completed, producing stress corrosion crack growth rate (CGR) reference curves for irradiated austenitic stainless steels in light water reactor (LWR) environments that were adopted as ASME B&PV Section XI Code Case N-889. As described in a technical basis Part 1 paper, over 800 CGR data points were collected from six laboratories worldwide, an international expert panel reviewed and ranked the data, and the better-ranked data were used to calibrate empirical models for irradiation-assisted stress corrosion cracking (IASCC) CGR in boiling water reactor (BWR) normal water chemistry (NWC), BWR hydrogen water chemistry (HWC) and pressurized water reactor (PWR) primary water environments. Part 1 also describes the custom fitting process, quality of fit, and comparisons with related literature and data not used for fitting. This technical basis Part 2 paper describes shifting the mean models to the 75th percentile of the calibration data, simplifying to produce the N-889 curves, and comparing with previous reference curves and over 500 data points not used for developing the N-889 curves, including weld, cast, and heat-affected-zone (HAZ) materials, additional wrought laboratory data, and field data from repeated inspection of BWR core shrouds. Part 2 also describes the irradiated yield stress model in Case N-889, compares that model with its calibration data and other data not used for calibration, and presents example calculations using both yield stress and CGR equations.

Technical Basis Part 1 for Code Case N-889: Data Selection and Modeling of Stress Corrosion Crack Growth Rates for Irradiated Austenitic Stainless Steels in Light Water Reactor Environments

Abstract A multiyear international data collection, data review, modeling, and implementation project was recently completed, producing stress corrosion crack growth rate (CGR) models and reference curves for irradiated austenitic stainless steels in light water reactor (LWR) environments. The curves have been approved as ASME B&PV Section XI Code Case N-889, and this paper is the technical basis Part 1 for Case N-889, summarizing collection, review, and selection of calibration data, modeling, and analysis of fit. Over 800 CGR data points were collected from six laboratories worldwide, an international expert panel reviewed and ranked the data, and the better-ranked data were used to calibrate empirical models for irradiation-assisted stress corrosion cracking (IASCC) CGR in boiling water reactor (BWR) normal water chemistry (NWC), BWR hydrogen water chemistry (HWC), and pressurized water reactor (PWR) primary water environments. This Part 1 paper also discusses related literature and compares with other data not used for mean model calibration. A technical basis Part 2 paper describes shifting the mean models to the 75th percentile, simplifying to produce Case N-889 reference curves, and comparing with previous reference curves and over 500 data points not used for reference curve development, including weld, cast, and heat affected zone (HAZ) materials, additional wrought laboratory data, and field data from repeated inspection of BWR core shrouds. Part 2 also describes the irradiated yield stress model in Case N-889, compares that model with its calibration data and other data not used for calibration, and presents example calculations.

Role of cavities on deformation-induced martensitic transformation pathways in a laser-welded, neutron irradiated austenitic stainless steel

The role of cavities on deformation-induced martensitic phase transformations is studied in a laser-welded and neutron irradiated austenitic stainless steel. Orientation dependent nanoindentation experiments are performed in the base metal and the weld heat affected zone (HAZ) at room temperature. Transmission electron microscopy study of deformed microstructures indicates indentation-induced α’-martensite forms in the base metal, whereas α’- and ε-martensite arise in the HAZ. The different pathway of martensite phase transformation is attributed to the laser weld-induced annealing of cavities. Our results suggest that deformation-induced martensitic phase transformation of austenitic stainless steel is correlated to neutron irradiated cavity structures.

Effect of laser welding on deformation mechanisms in irradiated austenitic stainless steel

Deformation mechanism of a laser weld on neutron irradiated AISI 304L stainless steel was studied by in-situ microcompression test at room temperature. The deformation-induced austenite-to-martensite phase transformation occurs in {101}-oriented grains in the irradiated base metal, while deformation twinning prevails in {101}-oriented grains in the weld heat affected zone (HAZ). A high number density of irradiation-induced voids in the base metal provide sufficient nucleation sites for the austenite-to-martensite phase transformation under compression at room temperature. A deformation map is established to predict critical twinning stress for face-centered cubic (fcc) metals and alloys. Our study suggests that irradiation can tailor the deformation-induced phase transformation in austenitic stainless steels.

Prediction of irradiation hardening in austenitic stainless steels: Analytical and crystal plasticity studies

The flow stress in irradiated Austenitic Stainless Steels (ASSs) is analyzed in terms of the microstructure features and the fundamental mechanisms of plastic deformation. A set of constitutive equations is proposed to assess the contributions of dislocation network, grain size effect, dislocation loops and solute clusters, accompanied with an appropriate superposition rule. Using simple equations for the evolution of densities of dislocations and radiation defects, it is possible to integrate numerically the tensile stress-strain curves over a wide range of irradiation doses. It is shown that irradiation hardening can be obtained analytically without adjustable parameters. Simulation results are challenged in comparison with experimental results and found in excellent agreement with all the reported experimental trends in ASSs irradiated at conditions close to those prevailing in pressurized water reactors.

Quantitative linkage between the stress at dislocation channel – Grain boundary interaction sites and irradiation assisted stress corrosion crack initiation

Localized deformation has emerged as a key factor in the crack initiation process for irradiated steels, as cracks are observed to nucleate preferentially at these sites. Using high resolution electron backscatter diffraction (HREBSD), the local stress tensor surrounding the dislocation channel-grain boundary interaction sites was quantified and coupled with fully determined grain boundary plane orientation information to determine, for the first time, the relationship between grain boundary normal stress and intergranular crack initiation in irradiated austenitic stainless steel. A Fe13Cr15Ni alloy was strained in simulated boiling water reactor, normal water chemistry after quantifying the residual stress tensor at discontinuous dislocation channel – grain boundary interaction sites where grain boundaries were determined to be well oriented with respect to the loading axis. Local stresses at the grain boundary were observed to reach magnitudes greater than 1.5 GPa at a distance of 200 nm from the intersection between the dislocation channel and the grain boundary. A pseudo-threshold stress of 0.9 GPa was measured, below which no cracking was observed. As the stress acting normal to the grain boundary increased above this value, the susceptibility to cracking increased with the cracking fraction reaching 100% at the high end of the stress range. This study shows for the first time that not only does intersection between discontinuous dislocation channels and grain boundaries result in peak local stresses, but the magnitude of the local tensile stress drives the crack initiation process.

Effect of Si-Addition in Stainless Steel on Surface Oxidation in Alkaline High Temperature Water

Model alloys 9Cr-25Ni-1.5Mo-5Si and 9Cr-25Ni-1.5Mo-3Si with lower Cr and higher Ni and Si contents than 316L SS were to simulate the grain boundary chemical compositions resulting from irradiation induced segregation in irradiated austenitic stainless steels. The effect of Si-addition in the model alloys on the oxide film formed in alkaline high temperature water at 290 oC was investigated by Raman spectroscopy, scanning electron microscopy and XPS. Compact layer was observed on all the alloys after exposure for 24h or longer. The outer layer oxides on model alloys were larger than that on 316L SS after the same immersion period. The oxide films formed on 316L SS and 9Cr-26Ni-5Si after exposure to the alkaline water at 290 oC for 120 h had a duplex structure. The oxide film formed on 316L SS was thicker than that on 9Cr-25Ni-1.5Mo-5Si alloy. Oxides of silicon was identified in the inner oxide layer 9Cr-25Ni-1.5Mo-5Si alloy but not significant on 316L SS.

Overview of Intergranular Fracture of Neutron Irradiated Austenitic Stainless Steels

Austenitic stainless steels are normally ductile and exhibit deep dimples on fracture surfaces. These steels can, however, exhibit brittle intergranular fracture under some circumstances. The occurrence of intergranular fracture in the irradiated steels is briefly reviewed based on limited literature data. The data are sorted according to the irradiation temperature. Intergranular fracture may occur in association with a high irradiation temperature and void swelling. At low irradiation temperature, the steels can exhibit intergranular fracture at low or even at room temperatures during loading in air and in high temperature water (~300 °C). This paper deals with the similarities and differences for IG fractures and discusses the mechanisms involved. The intergranular fracture occurrence at low temperatures might be correlated with decohesion or twinning and strain martensite transformation in local narrow areas around grain boundaries. The possibility of a ductile-to-brittle transition is also discussed. In case of void swelling higher than 3%, quasi-cleavage at low temperature might be expected as a consequence of ductile-to-brittle fracture changes with temperature. Any existence of the change in fracture behavior in the steels of present thermal reactor internals with increasing irradiation dose should be clearly proven or disproven. Further studies to clarify the mechanism are recommended.

Evolution of dislocation loops in austenitic stainless steels implanted with high concentration of hydrogen

It has been found that under certain conditions, hydrogen retention would be strongly enhanced in irradiated austenitic stainless steels. To investigate the effect of the retained hydrogen on the defect microstructure, AL-6XN stainless steel specimens were irradiated with low energy (100 keV) H2+ so that high concentration of hydrogen was injected into the specimens while considerable displacement damage dose (up to 7 dpa) was also achieved. Irradiation induced dislocation loops and voids were characterised by transmission electron microscopy. For specimens irradiated to 7 dpa at 290 °C, dislocation loops with high number density were found and the void swelling was observed. At 380 °C, most of dislocation loops were unfaulted and tangled at 7 dpa, and the void swellings were observed at 5 dpa and above. Combining the data from low dose in previous work to high dose, four stages of dislocation loops evolution with hydrogen retention were suggested. Finally, molecular dynamics simulation was made to elucidate the division of large dislocation loops under irradiation.

Effect of irradiation temperature on microstructural changes in self-ion irradiated austenitic stainless steel

We investigated the microstructural and hardness changes in austenitic stainless steel after Fe ion irradiation at 400, 300, and 200 °C using transmission electron microscopy (TEM) and nanoindentation. The size of the Frank loops increased and the density decreased with increasing irradiation temperature. Radiation-induced segregation (RIS) was detected across high-angle grain boundaries, and the degree of RIS increases with increasing irradiation temperature. Ni–Si clusters were observed using high-resolution TEM in the sample irradiated at 400 °C. The results of this work are compared with the literature data of self-ion and proton irradiation at comparable temperatures and damage levels on stainless steels with a similar material composition with this study. Despite the differences in dose rate, alloy composition and incident ion energy, the irradiation temperature dependence of RIS and the size and density of radiation defects followed the same trends, and were very comparable in magnitude.

Bayesian nonparametric analysis of crack growth rates in irradiated austenitic stainless steels in simulated BWR environments

To investigate influential parameters for irradiation-assisted stress corrosion cracking (IASCC) growth behavior, we attempt to analyze statistically existing data on the crack growth rate (CGR) in irradiated austenitic stainless steels (SSs) in boiling water reactor (BWR) environments using the Bayesian nonparametric (BNP) method. From the probability distribution of CGR and some input parameters, such as yield stress of irradiated material (σYS-irr), stress intensity factor (K), electrochemical corrosion potential (ECP), and fast neutron fluence, the mean CGR is estimated and compared with the measured CGR. The analytical results show good reproducibility of the measured CGR. The results also indicate the possible neutron fluence effects on CGR in high CGR region (i.e., high neutron fluence condition) by radiation-induced segregation (RIS), localized deformation, and/or other mechanisms than radiation hardening.

A study on the constitutive model of irradiated austenitic stainless steel for the functionality analysis of nuclear internals

The internal components of nuclear reactor are exposed to the neutron irradiation environment. The constituent materials of the components are susceptible to remarkable changes in the mechanical properties such as elastic modulus, strength, ductility and toughness. The mechanical and thermal deformations are accompanied with the void swelling and the irradiation creep in the environment. The functionality analysis which evaluates the structural integrity of the aged internals needs to take the degradation characteristics of the material into account. In this paper, a constitutive model of austenitic stainless steel developed by EPRI is studied and implemented into numerical analysis vehicles. The mechanical properties of irradiated 304 stainless steel are presented and the deformation behaviors are simulated. The criteria and methodology for the functionality analysis are also discussed and illustrated.

Influence of LBE long term exposure and simultaneous fast neutron irradiation on the mechanical properties of T91 and 316L

The LEXUR–II–LBE irradiation campaign was conducted from 2011 to 2012 and was aimed to investigate the combined influence of irradiation and LBE environment. In this irradiation campaign tensile test samples, pressurized tubes and corrosion samples were irradiated in LBE filled capsules. To separate the effect of exposure to LBE and neutron irradiation a parallel furnace experiment where the samples were exposed to LBE at the irradiation temperature for the corresponding time was conducted. Here we report results of the first extracted capsule which was irradiated about 6 months and dismantled after a cooling phase to decrease activity. The results of SSRT tests for irradiated T91 show that the exposure to LBE at 350 °C for a long time leads to the appearance of liquid metal embrittlement without any pre-treatment which is usually necessary to promote LME. Irradiation increases the effect of LME on the ductility of T91. In contrast to the findings for T91 the gained results also show that tensile tests on irradiated austenitic stainless steel 316L show no influence of LBE environment on the tensile properties.

Study on Magnetic Property Change on Neutron Irradiated Austenitic Stainless Steel

Study on Magnetic Property Change on Neutron Irradiated Austenitic Stainless Steel

Fatigue life predictions for irradiated stainless steels considering void swellings effects

The objective of this study is to estimate fatigue life of irradiated austenitic stainless steels types 304, 304L, and 316, which are extensively used as structural alloys in the internal elements of nuclear reactors. These reactor components are typically subjected to a long-term exposure to irradiation at elevated temperature along with repeated loadings during operation. Additionally, it is known that neutron irradiation can cause the formation and growth of microscopic defects or swellings in the materials, which may have a potential to deteriorate the mechanical properties of the materials. In this study, uniaxial fatigue models were used to predict fatigue properties based only on simple monotonic properties including ultimate tensile strength and Brinell hardness. Two existing models, the Bäumel–Seeger uniform material law and the Roessle–Fatemi hardness method, were employed and extended to include the effects of test temperature, neutron irradiation fluence, irradiation-induced helium and irradiation-induced swellings on fatigue life of austenitic stainless steels. The proposed models provided reasonable fatigue life predictions compared with the experimental data for all selected materials.