Irradiation-assisted Stress Corrosion Cracking Susceptibility Research Articles

Dislocation channel broadening–A new mechanism to improve irradiation-assisted stress corrosion cracking resistance of additively manufactured 316 L stainless steel

Additively manufactured (AM) 316 L stainless steel (SS) after hot isostatic pressing (HIP) was found to exhibit superior resistance to irradiation-assisted stress corrosion cracking (IASCC) in high-temperature water, as compared to wrought 316 L SS. The well-accepted IASCC factors of radiation-induced segregation (RIS) and radiation hardening are not accurate descriptions of IASCC susceptibility in this case. A decreased strain localization along grain boundaries, caused by dislocation channel broadening, was confirmed to suppress crack initiation. A unique distribution of irradiation-induced defects in HIP AM 316 L SS eased dislocation cross-slip compared to those in the wrought counterpart, thus increasing the channel width near the grain boundaries. For the first time, this study highlights the importance of dislocation channel broadening as a potential mechanism to further improve the IASCC resistance of 316 L SS and provides direct experimental evidence based on commercial-grade materials.

Correlation Between IASCC Susceptibility and Magnetic Properties of Stainless Steel Irradiated with Xenon Ions

In order to investigate the correlation between irradiation-assisted stress corrosion cracking (IASCC) susceptibility and the magnetic properties of austenitic stainless steel after ion irradiation, the magnetic properties and phase transformation were studied using a vibrating sample magnetometer and synchrotron radiation grazing incidence X-ray diffraction (GIXRD) on an irradiated specimen up to 25 displacements per atom. GIXRD indicated that the body-centered-cubic (BCC) phase was introduced in austenitic stainless steel by ion irradiation. Then, magnetic hysteresis loop analysis was conducted to give a quantitative analysis on the BCC magnetic phase under different irradiation damage and annealing temperature conditions. Based on the fact that postirradiation annealing will eliminate IASCC susceptibility at 550°C for 1 h, a correlation between IASCC susceptibility and magnetic properties was discussed, and a corresponding IASCC evaluation method was developed.

IASCC of neutron irradiated 316 stainless steel to 125 dpa

The irradiated microstructure and irradiation-assisted stress corrosion cracking (IASCC) susceptibility were determined for neutron-irradiated cold-worked (CW) 316 stainless steel tested in 320 °C simulated PWR primary water with various damage levels of 46.9, 67.4, and 125.4 dpa. The size and density of dislocation loops and precipitates were found to vary little over the dose range 46.9–125.4 dpa and dislocation loops were found to control the irradiation hardening. The %IG in constant extension rate tensile (CERT) tests and the stress to initiate a crack in four-point bend tests were consistent in establishing that the IASCC susceptibility was essentially unchanged across the dpa range. GB oxidization was observed prior to crack initiation, and localized deformation and triple junctions where local stresses were high were preferred crack initiation sites. IASCC cracks initiated preferentially on grain boundaries oriented nearly normal to the applied stress. An apparent threshold stress threshold for IASCC was determined to be ~45% of the irradiated yield strength for neutron-irradiated CW 316 SS between 46.9 and 125.4 dpa.

Microstructure, mechanical properties and IASCC susceptibility of stainless steel baffle bolts after 30 years of operation in a PWR

Neutron irradiation in the core of a nuclear power plant can change the material properties of stainless steel significantly. In this paper cold worked AISI 316 baffle bolts retrieved from Doel 1 PWR were investigated with the aim to assess the structural degradation due to neutron irradiation by analyzing microstructure, mechanical behavior and stress corrosion susceptibility. The baffle bolts had received neutron doses ranging from 6 to 59 dpa covering a major part of a nuclear power plant's lifetime. First, microstructural investigations were carried out to reveal the defect and dislocation density and the amount of void swelling. Then mechanical properties were determined using tensile tests and fracture toughness tests. The irradiation assisted stress corrosion cracking (IASCC) susceptibility was investigated by slow strain rate testing and crack growth rate testing. Results were compared with literature values and found to be in good agreement. Main conclusions were that void swelling was small even for the highest dpa values and that the fracture toughness values saturated above 15 dpa. The material was susceptible to IASCC as shown by SSRT and crack growth rate measurements exhibiting intergranular fracture morphology.

The effect of post-irradiation annealing on the stress corrosion crack growth rate of neutron-irradiated 304L stainless steel in boiling water reactor environment

The effect of post-irradiation annealing (PIA) on the irradiation-assisted stress corrosion cracking (IASCC) of 304L stainless steel in boiling water reactor environments was studied. PIA has little effect on the crack growth rate (CGR) in normal water chemistry even it restored most of the irradiation-induced microstructure change. Nevertheless, PIA can significantly suppress the CGR in hydrogen water chemistry (HWC). The mitigation factor of HWC increases with the degree of PIA up to 5 h at 550 °C and correlates with the recovery of Si segregation at grain boundary. PIA is an effective strategy for decreasing IASCC susceptibility in reducing water environment.

Effects of post-irradiation annealing on the IASCC susceptibility of neutron-irradiated 304L stainless steel

Post-irradiation annealing (PIA) was conducted on a 304L stainless steel irradiated to 5.9 dpa in the Barsebäck-1 BWR reactor, to investigate its effect on the mitigation of irradiation-assisted stress corrosion cracking (IASCC) susceptibility. IASCC susceptibility was measured for the as-irradiated and four PIA conditions (500 °C: 1 h and 550 °C: 1, 5, and 20 h) via interrupted constant extension rate tensile and four-point bend experiments under simulated BWR-NWC conditions. The annealing treatments were observed to progressively reduce IASCC susceptibility, as measured by the final intergranular fracture fraction (tensile) and crack length per unit area (four-point bend), with full removal of IASCC susceptibility being observed following annealing at 550 °C: 1 h for tensile specimens and 500 °C: 1h for four-point bend specimens. Among the microstructure and mechanical property parameters measured as a function of PIA, the average dislocation channel spacing was observed to decrease by ∼25% and ∼40% from the as-irradiated condition after annealing at 500 °C: 1 h and 550 °C: 1 h, respectively. The mitigation of IASCC susceptibility correlated well with the decrease in the average dislocation channel spacing and is consistent with a process in which crack initiation is controlled in part by the high tensile stress at dislocation channel-grain boundary intersections.

Radiation damage and irradiation-assisted stress corrosion cracking of additively manufactured 316L stainless steels

The irradiation-induced microstructure and irradiation-assisted stress corrosion cracking (IASCC) behavior of additively manufactured (AM) 316L stainless steels produced by laser powder bed fusion were evaluated for the first time. Irradiation-induced dislocation loops, voids, and γ′ precipitates were observed in all processing conditions following 2.5 dpa at 360 °C. The cell structure and dense dislocation walls in the stress-relieved AM materials recovered and showed signs of recrystallization following irradiation. Anisotropy in both tensile property and IASCC susceptibility were observed in the stress-relieved AM 316L stainless steel due to the printing texture. The hot-isotropic pressed AM 316L had better radiation tolerance and lower IASCC susceptibility than the stress-relieved AM 316L and conventionally forged 316L. Therefore, post-printing hot-isotropic pressing (HIP) is recommended for enhancing radiation tolerance and IASCC performance in nuclear applications as it eliminates the anisotropic mechanical behavior and IASCC susceptibility associated with the printing texture.

Helium effects on IASCC susceptibility in as-implanted solution annealed, cold-worked and post-implantation annealed 316L steel

ABSTRACTThe susceptibility to irradiation-assisted stress corrosion cracking (IASCC) at relatively low dose is dominated by conventional mechanisms such as radiation-induced segregation, hardening, creep/relaxation and radiolysis. Large amount of helium (He) can be accumulated in reactor internal components made of nickel containing steels of pressurised water reactors after long-term operation. The increase of He content has been correlated to IASCC susceptibility in reactor internals subjected to neutron irradiation. However, the question whether He increases or even dominates the IASCC susceptibility at high doses is still open. In this work, He was homogeneously implanted at 300°C to 300 and 1000 appm. Slow strain rate tests were performed in as-implanted solution annealed and cold-worked samples and in post-implantation annealed samples. The results show that homogenised implanted He up to 1000 appm alone (without significant displacement damage) does not produce intergranular cracking and IASCC in high-temperature air and in hydrogenated high-temperature water, respectively.

Crack initiation behavior of neutron irradiated model and commercial stainless steels in high temperature water

The objective of this study was to isolate key factors affecting the irradiation-assisted stress corrosion cracking (IASCC) susceptibility of eleven neutron-irradiated austenitic stainless steel alloys. Four commercial purity and seven high purity stainless steels were fabricated with specific changes in composition and microstructure, and irradiated in a fast reactor spectrum at 320°C to doses between 4.4 and 47.5dpa. Constant extension rate tensile (CERT) tests were performed in normal water chemistry (NWC), hydrogen water chemistry (HWC), or primary water (PW) environments to isolate the effects of environment, elemental solute addition, alloy purity, alloy heat, alloy type, cold work, and irradiation dose. The irradiated alloys showed a wide variation in IASCC susceptibility, as measured by the relative changes in mechanical properties and crack morphology. Cracking susceptibility measured by %IG was enhanced in oxidizing environments, although testing in the lowest potential environment caused an increase in surface crack density. Alloys containing solute addition of Ni or Ni+Cr exhibited no IASCC. Susceptibility was reduced in materials cold worked prior to irradiation, and increased with increasing irradiation dose. Irradiation-induced hardening was accounted for by the dislocation loop microstructure, however no relation between crack initiation and radiation hardening was found.

Development of a magnetic sensor system for predictive IASCC diagnosis on stainless steels in a nuclear reactor

The authors previously reported that magnetic flux leakage d ata, measured using a flux gate (FG) sensor, showed a correlation with the irradiation-assisted stress corros ion cracking (IASCC) susceptibility of neutron-irradiate d austenitic stainless alloys. This paper presents a study conducted to develop a diagnostic system that can detect IASCC precursors in stainless steels by measuring the magnetic properties of the material. The eddy current method and alternating current (AC) magnetization method were used, as these will be more practical for use in actual reactors. Probes were developed for these measurement methods, providing sufficient tolerance for en vironments in nuclear reactors. An attempt was also made to improve spatial resolution by manufacturing a smaller probe. A sensor system was designed for remote control, performance tests were conducted by measuring neutron-irradiated specimens and mock specimens, and magnetic data were evaluated by comparing the IASCC susceptibility of the specimens. It was proved that the sensor system developed in this study is capable of detecting IASCC precursors. Further developments necessary for application in actual nuclear reactors and the mechanism of correlation between magnetic properties and IASCC susceptibility were also discussed. Many of the problems experienced in existing light water reactors (LWR) are caused by damage to structural materials. Stress corrosion cracking (SCC) and irradiation-assisted SCC (IASCC) of structural materials have been especially serious problems, and it has not been possible to predict these in the reactor design process. These problems will become more significant when existing reactors start to age from now on, and when advanced reactors are operated and structural materials are used in more severe irradiation environments in future. It is therefore import ant to develop diagnostic techniques that can

Radiation hardening and IASCC susceptibility of extra high purity austenitic stainless steel

An extra high purity austenitic stainless steel (EHP alloy: Fe–25Cr–35Ni SAR) was developed with countermeasures for irradiation assisted stress corrosion cracking (IASCC). To study the radiation-induced hardening and the susceptibility of IASCC for EHP alloy, EHP alloys were irradiated at 553±20K in a helium and nitrogen gas mixture at the fluence level of 1.5×1025n/m2. After neutron irradiation, tensile tests and slow strain rate tests (SSRTs) of irradiated specimens were conducted in air at 561K and in oxygenated high purity water at 561K, respectively. The fracture surfaces were examined using SEM after tensile tests and SSRTs. The tendency of radiation-induced hardening of EHP alloy was the same as irradiated type 304 and type 316 stainless steels. Though there was a little effect of the countermeasures on the suppression of irradiation hardening. EHP alloy had excellent resistance for IASCC up to 1.5×1025n/m2.

Localized deformation and IASCC initiation in austenitic stainless steels

Localized deformation may play a key role in the underlying mechanism of irradiation assisted stress corrosion cracking (IASCC) in light water reactor core components. In this study, four austenitic alloys, 18Cr8Ni, 15Cr12Ni, 13Cr15Ni and 21Cr32Ni, with different stacking fault energies were irradiated to 1 and 5 dpa at 360 °C using 3.2 MeV protons. Interrupted constant extension rate tensile (CERT) tests were conducted in a simulated BWR environment to determine IASCC susceptibility. In order to characterize the localized deformation in slip channels and grain boundaries, parallel CERT experiments were also performed in an argon atmosphere. Results show that the IASCC susceptibility of the tested alloys increases with increasing irradiation dose and decreasing stacking fault energy. IASCC tends to initiate at locations where slip channels intersect grain boundaries. Localized deformation in the form of grain boundary sliding due to the interaction of slip channels and grain boundaries is likely the primary cause of the observed cracking initiation.

Deformation microstructure of proton-irradiated stainless steels

The deformation microstructure of proton-irradiated stainless steels may play a key role in explaining their irradiation-assisted stress corrosion cracking (IASCC) susceptibility. In the present study, three model alloys (UHP-304, 304+Si, 304+Cr+Ni) with different stacking fault energies (SFEs) were irradiated with 3.2MeV protons at 360°C to 1.0 and 5.5dpa and then strained in 288°C Ar atmosphere. The deformation microstructure of the strained samples was investigated using scanning electron microscopy and transmission electron microscopy. The results showed that the slip lines interacted with grain boundaries by grain-to-grain transmission, grain boundary sliding or deformation ledge formation at grain boundaries. Expanded channels, which were formed at locations where dislocation channels intersected the grain boundaries or other channels, were found predominately in the low SFE alloys UHP-304 and 304+Si. The steps and shear strain at grain boundaries caused by channel expansion may increase the IASCC susceptibility in low SFE stainless steels by producing strain concentrations and inducing cracks in the oxide film.

Separation of Microstructural and Microchemical Effects in Irradiation Assisted Stress Corrosion Cracking using Post-irradiation Annealing

Post-irradiation annealing (PIA) was conducted in order to clarify the role of microstructural and microchemical effects on irradiation assisted stress corrosion cracking (IASCC) susceptibility in simulated pressurized water reactor (PWR) primary water. Microstructures, hardening, radiation-induced segregation and IASCC susceptibility were examined in cold-worked SUS 316 stainless steels irradiated to 25 dpa in a PWR after annealing at 673–823 K for 1h. IASCC susceptibility, microstructures and hardening recovered as the annealing temperature increased whereas the grain boundary segregation of Cr and Ni remained almost unchanged. The results suggested that the change in IASCC susceptibility due to annealing is not attributed to the change in grain boundary segregation but to the change in micro-structures and hardening. The fact that a smaller recovery of radiation hardening caused a larger IASCC susceptibility suggested that a threshold hardening level exists for the occurrence of IASCC.

Effects of water and irradiation temperatures on IASCC susceptibility of type 316 stainless steel

Effects of water and irradiation temperatures on irradiation-assisted stress corrosion cracking (IASCC) of type 316 stainless steel were investigated. Type 316 stainless steel was irradiated at 333–673 K to a dose level of 16 dpa. Susceptibility to IASCC was evaluated by slow strain rate testing in oxygenated water in the temperature range of 513–573 K. Irradiation at 603 and 673 K caused IASCC in 513 K water, but irradiation below 473 K did not induce IASCC at 513 K. Specimens irradiated at 333 K did not show IASCC susceptibility in 513 K water, but high susceptibility was observed in 573 K water. Effect of irradiation temperature is discussed from the view points of microstructural and microcompositional changes.

IASCC Susceptibility and Slow Tensile Properties of Highly-irradiated 316 Stainless Steels

Irradiation assisted stress corrosion cracking (IASCC) of cold-worked 316 stainless steels irradiated to doses up to 53 dpa was examined using slow strain rate tensile tests in 593 K simulated pressurized water reactor primary water while changing the content of dissolved hydrogen (DH) and dissolved oxygen (DO). A higher susceptibility was observed for higher doses and DH content, accompanied by increased corrosion product formation on the fracture surface and higher hydrogen accumulation near the fracture surface. At 53 dpa the susceptibility at both 0.02 and 8ppm DO was comparable to that at high DH content. The results indicated that IASCC was sensitive to DH content at doses less than 35 dpa but was less sensitive to both DH and DO content at 53 dpa. The subcrack formation and hydrogen accumulation in the hydrogenated condition suggested that processes associated with hydrogen would have an important role in IASCC in hydrogenated water. The same stainless steels were susceptible to intergranular type fracture during slow tensile tests in pure argon at 593 K. The intergranular type region consisted of a mixture of intergranular and dimple regions and the intergranular fraction was much smaller than that in IASCC in water environment.

IASCC Susceptibility and Slow Tensile Properties of Highly-irradiated 316 Stainless Steels

Irradiation assisted stress corrosion cracking (IASCC) of cold-worked 316 stainless steels irradiated to doses up to 53 dpa was examined using slow strain rate tensile tests in 593 K simulated pressurized water reactor primary water while changing the content of dissolved hydrogen (DH) and dissolved oxygen (DO). A higher susceptibility was observed for higher doses and DH content, accompanied by increased corrosion product formation on the fracture surface and higher hydrogen accumulation near the fracture surface. At 53 dpa the susceptibility at both 0.02 and 8ppm DO was comparable to that at high DH content. The results indicated that IASCC was sensitive to DH content at doses less than 35 dpa but was less sensitive to both DH and DO content at 53 dpa. The subcrack formation and hydrogen accumulation in the hydrogenated condition suggested that processes associated with hydrogen would have an important role in IASCC in hydrogenated water. The same stainless steels were susceptible to intergranular type fracture during slow tensile tests in pure argon at 593 K. The intergranular type region consisted of a mixture of intergranular and dimple regions and the intergranular fraction was much smaller than that in IASCC in water environment.

Irradiation-assisted SCC susceptibility of HIPed 316LN-IG stainless steel irradiated at 473 K to 1 dpa

Solid hot-isostatic pressed (solid-HIPed) joint specimens and those with or without thermal-cycled specimens of 316LN-IG were irradiated at about 473 K to 1 dpa in the high flux isotope reactor. Slow strain rate tests were conducted in a high-purity, oxygenated ( dissolved oxygen =10 wt ppm) water at 423, 513 and 573 K with strain rates of (2–10)×10 −7 s −1. Tensile tests were also conducted in vacuum at the same temperature range. Fracture surfaces were observed by scanning electron microscopy. No specimen showed irradiation-assisted stress corrosion cracking (IASCC) susceptibility at 423 and 513 K in water. At 573 K, however, intergranular cracks were observed to form in HIPed specimens. It was concluded that the effect of HIPing to IASCC susceptibility is small.

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.

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.