J. T. Busby

Emulation of neutron irradiation effects with protons: validation of principle

Abstract 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.

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.

On the mechanism of radiation-induced segregation in austenitic Fe-Cr-Ni alloys

Abstract The relative importance of the vacancy and interstitial contributions to radiation-induced segregation (RIS) in Fe–Cr–Ni alloys is studied to better understand the mechanisms causing changes in grain boundary composition and to improve the capability to predict RIS in austenitic Fe–Cr–Ni alloys. The primary driving mechanism for segregation in Fe–Cr–Ni alloys is shown to be the inverse Kirkendall (IK) mechanism, specifically the coupling between alloying elements and the vacancy flux. To study grain boundary segregation, seven alloys were irradiated with 3.2 MeV protons at temperatures from 200°C to 600°C and to doses from 0.1 to 3 dpa. Grain boundary compositions were measured using both Auger electron spectroscopy (AES) and scanning transmission electron microscopy with energy dispersive X-ray spectroscopy (STEM/EDS). Grain boundary compositions were compared to model predictions that assume segregation was driven either by preferential interaction of solute atoms with the vacancy flux alone or in combination with binding of undersized solutes to the interstitial flux. Calculations that assume the segregation is caused by preferential interaction of solute atoms with the vacancy flux generally followed the trends of the segregation measurements. However, the inclusion of interstitial binding to the IK model causes poor agreement between model predictions and segregation measurements, resulting in severe overprediction of Ni enrichment and Fe depletion. Comparisons of segregation models with RIS in alloys irradiated with neutrons also show that preferential interaction of solutes with the vacancy flux sufficiently describes segregation in Fe–Cr–Ni alloys.

Effect of proton and Ne irradiation on the microstructure of Zircaloy 4

Proton irradiation at temperatures of 310 and 350°C has been explored as a method of studying radiation effects in Zircaloy 4 under power-reactor conditions. The dislocation-loop and hardness evolution are similar to those observed under neutron irradiation. Amorphization and Fe loss at Zr(Cr,Fe)2 precipitates are observed. Energy-filtered images indicate Fe accumulation in the matrix near the precipitate, a phenomenon not reported in neutron-irradiated alloys.

Microchemistry and microstructure of proton-irradiated austenitic alloys : toward an understanding of irradiation effects in LWR core components

Abstract Over 1200 measurements of grain boundary composition and microstructure have been made on 14 different austenitic Fe–Cr–Ni alloys following proton irradiation in the temperature range 200–600°C and in the dose range 0.1–3.0 dpa. From these data, a greater understanding of radiation induced segregation (RIS) and microstructure development has been gained. Grain boundary composition measurements revealed that Cr depletes at grain boundaries, Ni enriches and Fe can either enrich or deplete depending on alloy composition. Analysis of temperature and composition dependence of RIS revealed that the magnitude and direction of grain boundary segregation depends on alloy composition because the values of migration enthalpy of the alloy constituents are not the same, and diffusivities of the alloy constituents are composition-dependent. The dose dependence of segregation revealed ordering in Ni-base alloys and temperature dependence was used to show that RIS is consistent with a vacancy exchange mechanism. The dependence of segregation on composition is consistent with all known, relevant neutron data. RIS was found to be related to the development of the dislocation and void microstructures. Alloys in which the microstructure develops slower with dose also show slower changes in RIS. Similarly, it was shown that the dependence of swelling on composition is the same for neutron, ion and proton irradiation and all can be explained by the effect of RIS on defect diffusivity at the void nuclei. This paper illustrates the value of conducting carefully chosen irradiation experiments over several, well-controlled variables to elucidate the mechanisms underlying the microchemical and microstructural changes.

Role of irradiated microstructure and microchemistry in irradiation-assisted stress corrosion cracking

Irradiation has a profound effect on the stress corrosion cracking propensity of austenitic alloys in high-temperature water. Irradiation-assisted stress corrosion cracking (IASCC) has been well documented both in the laboratory and in service over the past two decades. Numerous studies have shown that the degree of intergranular stress corrosion cracking increases with dose. However, the microstructure is simultaneously changing in several ways (dislocation loops, voids, segregation and hardening) and, not surprisingly, they all correlate with increased cracking susceptibility. As a consequence of their simultaneous development, the attribution of IASCC to one or more of these features has been difficult to establish. Mechanisms based on each of the principal effects of irradiation in the alloy are considered. Arguments can be made in support of any one of these features as the cause of IASCC, but substantial evidence exists to refute a first order correlation. The mechanism of IASCC is more likely due to a combination of factors, or second order effects not yet considered. One such mechanism that is considered is based on the change in the deformation mode caused by the irradiated microstructure and the interaction of localized deformation bands with grain boundaries.

Localized deformation induced IGSCC and IASCC of austenitic alloys in high temperature water

Grain boundary properties are known to affect the intergranular stress corrosion cracking (IGSCC) and irradiation assisted stress corrosion cracking behavior of austenitic alloys in high temperature water. However, it is only recently that sufficient evidence has accumulated to show that the disposition of deformation in and near the grain boundary plays a key role in intergranular cracking. Grain boundaries that can transmit strain to adjacent grains can relieve stresses without undergoing localized deformation. Grain boundaries that cannot transmit strain will either experience high stresses or high strains. High stresses can lead to wedge-type cracking and sliding can lead to rupture of the protective oxide film. These processes are also applicable to irradiated materials in which the deformation can become highly localized in the form of dislocation channels and deformation twins. These deformation bands conduct tremendous amounts of strain to the grain boundaries. The capability of a boundary to transmit strain to a neighboring grain will determine its propensity for cracking, analogous to that in unirradiated metals. Thus, IGSCC in unirradiated materials and IASCC in irradiated materials are governed by the same local processes of stress and strain accommodation at the boundary.

Post-irradiation annealing of small defect clusters

Post-irradiation annealing studies indicate small defect clusters may be a potential contributor to IASCC. In this study, small defect clusters and their behavior during annealing are examined. A CP-304 SS alloy was irradiated with 3.2u2009MeV protons to 0.17u2009dpa at <75°C. The increase in hardness after 0.17u2009dpa at <75°C was greater than that after 0.55u2009dpa at 360°C, but dropped significantly after annealing at 350°C for short times, although, no dislocation damage was observed using transmission electron microscopy (TEM). For samples irradiated in a previous study at 360°C, the smallest dislocation loops and clusters were removed preferentially during annealing at 500°C for 45 minutes. Computer simulations of the annealing behavior were performed for the data from both sets of experiments in order to gain further insight into the nature of small clusters.

The Correlation Between Swelling and Radiation-Induced Segregation in Iron-Chromium-Nickel Alloys

The magnitudes of both void swelling and radiation-induced segregation (RIS) in iron-chromium-nickel alloys are dependent on bulk alloy composition. Because the diffusivity of nickel via the vacancy flux is slow relative to chromium, nickel enriches and chromium depletes at void surfaces during irradiation. This local composition change reduces the subsequent vacancy flux to the void, thereby reducing void swelling. In this work, the resistance to swelling from major element segregation is estimated using diffusivities derived from grain boundary segregation measurements in irradiated iron-chromium-nickel alloys. The resistance to void swelling in iron- and nickel-base alloys correlates with the segregation and both are functions of bulk alloy composition. Alloys that display the greatest amount of nickel enrichment and chromium depletion are found to be most resistant to void swelling, as predicted. Additionally, swelling is shown to be greater in alloys in which the RIS profiles are slow to develop.

Microchemistry of proton-irradiated austenitic alloys under conditions relevant to LWR core components

Over 1200 measurements of grain boundary composition and microstructure have been made on 14 different austenitic Fe-Cr-Ni alloys following proton irradiation in the temperature range 200-600°C and in the dose range 0.1-3.0 dpa. Grain boundary composition measurements revealed that Cr depletes at grain boundaries, Ni enriches and Fe can either enrich or deplete depending on alloy composition. Analysis of temperature and composition dependence of RIS revealed that the magnitude and direction of grain boundary segregation depends on alloy composition because the value of migration enthalpy differs among the alloy constituents, and diffusivities of the alloy constituents are composition-dependent. The dose dependence of segregation revealed ordering in Ni-base alloys and temperature dependence was used to show that RIS occurs by vacancy exchange rather than an interstitial binding mechanism. The dependence of segregation on composition is consistent with all known, relevant neutron data.