Proton-irradiated Austenitic Stainless Steels Research Articles

Proton irradiation for radiation-induced changes in microstructures and mechanical properties of austenitic stainless steel

We report on microstructural and mechanical property changes as a function of radiation damage value in proton-irradiated austenitic stainless steel by means of advanced characterization techniques. The microstructural changes in proton-irradiated austenitic stainless steel were analyzed by transmission electron microscopy for observation of radiation-induced defects as well as the measurement of the chemical composition at grain boundaries. The radiation hardening after the proton irradiation was characterized by nano indentation for changes in hardness profiles with radiation damage.Various transition points for microstructural and mechanical property changes under proton irradiation are analyzed via material characterization of proton-irradiated austenitic stainless steels. The saturation is expected to occur at approximately 10 displacements per atom (dpa) for the radiation-induced segregation of Cr, Ni, and P and approximately 2.5 dpa for radiation hardening. The cavity formation is observed to occur at hydrogen concentration levels greater than 5E5 atomic parts per million (appm) H. It is also found that the transition from black dot to Frank loop happened above approximately 1 dpa.Profiles of radiation-induced segregation and radiation hardening as a function of dpa can be extended to the high irradiation condition, and can be compared with experimental data for neutron irradiation-induced segregation and radiation hardening. The radiation-induced segregation after the proton irradiation at 360 °C are in good agreement with that after neutron irradiation. On the other hand, it is observed that the evolution of radiation-induced defects and the corresponding radiation hardening exhibit sooner, that appears to be because of the dose rate effect.

Evaluation of critical resolved shear strength and deformation mode in proton-irradiated austenitic stainless steel using micro-compression tests

Micro-compression tests were applied to evaluate the changes in the strength and deformation mode of proton-irradiated commercial austenitic stainless steel. Proton irradiation generated small dots at low dose levels and Frank loops at high dose levels. The increase in critical resolved shear stresses (CRSS) was measured from micro-compression of pillars and the Schmid factor calculated from the measured loading direction. The magnitudes of the CRSS increase were in good agreement with the values calculated from the barrier hardening model using the measured size and density of radiation defects. The deformation mode changed upon increasing the irradiation dose level. At a low radiation dose level, work hardening and smooth flow behavior were observed. Increasing the dose level resulted in the flow behavior changing to a distinct heterogeneous flow, yielding a few large strain bursts in the stress–strain curves. The change in the deformation mode was related to the formation and propagation of defect-free slip bands. The effect of the orientation of the pillar or loading direction on the strengths is discussed.

Effect of prior cold-work on radiation-induced segregation in proton-irradiated austenitic stainless steel

The effect of prior cold-work on radiation-induced segregation in proton-irradiated type 304 SS (1.00dpa) was investigated using electrochemical potentiokinetic reactivation (EPR) test followed by atomic force microscopic examination. Attacked linear features were noticed after EPR testing of the irradiated specimen and it was attributed to chromium depletion and/or decoration of strain-regions by point-defects on slip planes. Attack on locations near grain boundaries was also noticed after EPR tests in the irradiated specimen. Electron backscatter diffraction analysis of the irradiated specimen showed increase in Σ1 fraction within the matrix and formation of strain-regions near grain boundaries.

Fatigue Crack Initiation in Proton-Irradiated Austenitic Stainless Steel

The effect of irradiation on slip band formation and growth and microcrack initiation behavior under low cycle fatigue in SUS316L austenitic stainless steel was investigated using accelerator-based proton irradiation and a low cycle fatigue test at room temperature in air. The mean space of the slip line in proton-irradiated specimens was 25–40% wider than that in unirradiated specimens under the same number of cycles, possibly due to localized deformation by proton irradiation. The microcrack initiation life of the proton-irradiated specimens was approximately 20% of that of the unirradiated specimens. While the microcrack initiation in the unirradiated specimens was observed at the grain boundary, twin boundary, slip band, and triple junction, that in the proton-irradiated specimens was observed only at the twin boundary and slip band, possibly due to irradiation hardening. The step-height of an extrusion near the microcrack was almost the same in the unirradiated and proton-irradiated specimens regardless of the initiation site (100–150 nm). Therefore, the microcrack initiation was considered to occur when the surface morphology change involving the extrusion exceeded the specific threshold value.

Fatigue Crack Initiation in Proton-Irradiated Austenitic Stainless Steel

Fatigue Crack Initiation in Proton-Irradiated Austenitic Stainless Steel

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.

Microstructural aspect of irradiation assisted stress corrosion cracking of LWR in-core structural materials

Microstructural evolution of proton-irradiated austenitic stainless steels and Inconel 600 alloy were studied in the temperature range interesting to light water reactor (LWR) core conditions. Radiation-induced solute segregation was observed in all irradiation temperatures at dose of 0.1 dpa and above. Constant extension rate test (CERT) was performed to study the irradiation-assisted stress corrosion cracking (IASCC) behavior of these alloys. Void swelling, dislocation and loop structures, and phase transformation were studied to understand the relationship of microstructural evolution to the IASCC phenomenon. It is found that in low carbon austenitic stainless steels, the γ → α phase transformation could play an important role in stress corrosion cracking mechanism. Twining dislocations were observed in 304L specimens irradiated at 230°C to 1 dpa.

Deformation Mechanisms in a Proton-Irradiated Austenitic Stainless Steel

AbstractSamples of austenitic 304L stainless steel have been irradiated with 3.4 MeV protons to atotal dose of 1 dpa. The microstructure of the irradiated stainless steel has been quantified by transmission electron microscopy and shown to be similar to that found in neutron-irradiated core components. Constant extension rate tensile tests have been performed at strain rates of 3x10−7 s−1 and 3x10−8 s−1 to strains of up to 27% at 23°C and 288°C. The resulting microstructures were characterized using electron and optical microscopy. Deformation of the unirradiated material is similar to that reported by others in previous work on austenitic steels, consisting of dislocation source activation and formation of dense dislocation networks with increasing strain. In the irradiated samples tested at 288°C, deformation consists of dislocation source activation at grain boundaries punching dislocations through the grain interior to the opposing grain boundaries. The dislocations create channels that are free of radiation-produced defects and on which dislocation motion is concentrated. Dislocations in the channels pile-up at the boundaries, creating regions of highly localized stress concentration at the grain boundary. The mechanism by which this stress is relieved is still unknown. Deformation at 23°C consists of the nucleation and propagation of microtwins across the width of the grains.

Effect of additional elements on cavity formation near grain boundary in proton-irradiated austenitic stainless steels

Using 140 keV H 2 + ion irradiation, the segregation of hydrogen along grain boundaries in type 304L, Ti- and Nb-modified austenitic stainless steels was studied. The irradiation was performed at temperatures up to 673 K and doses up to 3 × 10 21 H 2 +/m 2. Cavities formed along grain boundaries in type 304L after irradiation. The number density and diameter of the cavities were less in the Ti- and Nb-modified steels compared with that of 304L. It appears that Ti and Nb contribute to a decrease in the amount of hydrogen trapped along grain boundaries.

Effects of dose rate on void formation and precipitation in proton irradiated austenitic stainless steels

The effects of dose rate on void swelling and precipitation behaviour during irradiation were examined in type 316 stainless steel and the Japanese prime candidate alloy for a fusion reactor first wall (J-PCA). The specimens were irradiated at 823 K by 200 keV protons to 20 dpa at a dose rate of 2.0 × 10 −4dpa/s for 0.1 Ms (28 h) and 1.5 × 10 −5 dpa/s for 1.37 Ms (380 h), respectively. Notable differences in swelling and precipitation behaviour were not observed between the high and low dose rate irradiations in type 316 stainless steel. In the J-PCA, on the other hand, the low dose rate irradiation showed much lower swelling and a much greater amount of intragranular TiC precipitation than the high dose rate irradiation for both solution-annealed and cold-worked conditions.